Text

Long-Term Containment Analysis, Revision 2
	ML032721450
Person / Time
Site:	Monticello
Issue date:	08/31/2003
From:	Rhow S; General Electric Co
To:	; Office of Nuclear Reactor Regulation
References
	L-MT-03-067, TAC MB7185 GE-NE-0000-0002-8817-01, Rev 2
	Download: ML032721450 (148)
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Attachment 3 NUCLEAR MANAGEMENT COMPANY, LLC MONTICELLO NUCLEAR GENERATING PLANT DOCKET 50-263 GE NUCLEAR ENERGY REPORT, GE-NE-0000-0002-8817-01-R2 MONTICELLO NUCLEAR GENERATING PLANT LONG-TERM CONTAINMENT ANALYSIS, REVISION 2, DATED AUGUST 2003, NON-PROPRIETARY VERSION GE Report follows

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I1, GE Nuclear Energy GE-NE-0000-0002-8817-01 Revision 2 Class I August 2003 Monticello Nuclear Generating Plant Long-term Containment Analysis

GE -NE-000-0002-8817-01 Revision 2 Class I August 2003 Monticello Nuclear Generating Plant Long-term Containment Analysis Sang K. Rhow Approval:

Israel Nir, Manager Nuclear and Safety Analysis

GE-NE-0000-0002-8817-01-R2 REVISION The following changes to the Revision 1 report, GE-NE-0000-0002-8817-01-RI, September 2002, are incorporated in this Revision 2 report

The initial DW temperature listed in Tables 3-2 and 3-3 is changed from 1500F to 1350F.

In I' sentence of 2nd paragraph, page 4-17, the peak suppression pool temperature and its timing are changed from 195.80F and 37163 seconds to 196.50F and 37104 seconds, respectively.

In 2nd sentence of 2nd paragraph, page 4-17, the long-term peak suppression chamber pressure is changed from 18.8 psig to 19 psig.

In ' sentence of 2d paragraph, page 4-18, the peak suppression pool temperature is changed from 196.50F to 196.2 0F.

In " sentence of 2d paragraph, page 4-18, the peak suppression pool temperature is changed from 195.80F to 196.50F.

In 2nd sentence of 2nd paragraph, page 4-18, the word "higher" is changed to "slightly lower."

GE-NE-0000-0002-8817-01-R2 IMPORTANT NOTICE REGARDING THE CONTENTS OF THIS REPORT Please Read Carefully A. Disclaimer The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between the company receiving this document and GE. Nothing contained in this document shall be construed as changing the applicable contract. The use of this information by anyone other than a customer authorized by GE to have this document, or for any purpose other than that for which it is intended, is not authorized. With respect to any unauthorized use, GE makes no representation or warranty, and assumes no liability as to the completeness, accuracy or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

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GE-NE0000-0002-8817-01-R2 TABLE OF CONTENTS Page ACRONYMS AND ABBREVIATIONS..............................

.......................... vi

1. PROJECT OVERVIEW AND OBJECTIVES........................................................ 1-1

2.

SUMMARY

AND CONCLUSIONS........................................................

2-1 2.1 Summary........................................................

2-1 2.2 Conclusions........................................................

2-3

3. EVALUATION........................................................

3-1 3.1 Method of Evaluation........................................................

3-1 3.2 Inputs and Assumptions for SAFER/GESTR Analysis of Isolation Event and Small Breaks........................................................

3-5 3.3 Inputs and Assumptions for SHEX Containment Analysis........

.................... 3-8 3.3.1 DBA-LOCA with Direct Suppression Pool Cooling............................ 3-8 3.3.2 DBA-LOCA with Containment Spray Cooling for NPSH..........

....... 3-11 3.3.3 Reactor Isolation and Small Breaks with Direct Suppression Pool Cooling................................................................................................ 3-12

4. EVALUATION RESULTS 4-1 4.1 ECCS-LOCA Results for Isolation Event and Small Breaks without HPCI.. 4-1 4.2 Containment Analysis Results for Isolation Event, Small Breaks and DBA-LOCA with 143.1 Btu/sec-0F K-Value and 900F Service Water..................

4-4 4.2.1 Isolation Event, Small-break LOCAs and DBA-LOCA with One RHR Loop.....................................................

4-5 4.2.2 DBA-LOCA as Limiting Event.....................................................

4-8 4.2.3 Isolation Event with Two RHR Loops................................................... 4-8 4.2.4 Containment Response to LOCAs with Operable HPCI............

........... 4-9 4.3 DBA-LOCA Analysis for Short-term NPSH Evaluation.

4-11 4.4 DBA-LOCA with 147 Btu/sec-0F K-Value and 900F Service Water...........

4-12 4.4.1 DBA-LOCA with Direct Suppression Pool Cooling.......................... 4-12 4.4.2 DBA-LOCA with Containment Spray Cooling for NPSH...........

....... 4-13 ii

GE-NE-0000-0002-8817-01-R2 4.4.3 Impact of 60-second Delay in Realignment to Containment Cooling Mode....................................................

4-14 4.5 DBA-LOCA with 147 Btu/sec-0F K-Value and 940F Service Water........... 4-15 4.5.1 DBA-LOCA with Direct Suppression Pool Cooling.......................... 4-16 4.5.2 DBA-LOCA with Containment Spray Cooling for NPSH.................. 4-17

5. REFERENCES....................................................

5-1 APPENDIX A: CORE HEAT VALUES USED IN CONTAINMENT ANALYSIS....................................................

A-1 APPENDIX B: SHEX BENCHMARKING ANALYSIS FOR DBA-LOCA............. B-1 APPENDIX C: TABLE FOR SHORT-TERM CONTAINMENT RESPONSE TO DBA-LOCA FOR INPUT TO NPSH..................................................

C-1 APPENDIX D: TABLES FOR LONG-TERM CONTAINMENT RESPONSE TO DBA-LOCA FOR INPUT TO NPSH (K=147, SWT=900F)......................... D-1 APPENDIX E: TABLES FOR LONG-TERM CONTAINMENT RESPONSE TO DBA-LOCA FOR INPUT TO NPSH (K=147, SWT-940F)................................ E-1 APPENDIX F: PLOTS FROM SAFER/GESTR RUNS............................................. F-1 APPENDIX G: PLOTS FROM SHEX RUNS FOR ISOLATION EVENTS, SMALL BREAKS AND DBA-LOCA WITH DIRECT POOL COOLING (K=143.1, SWT-900F).................................................

G-APPENDIX H: PLOTS FROM SHEX RUN FOR SHORT-TERM NPSH DBA-LOCA.................................................

H-1 APPENDIX I: PLOTS FROM SHEX RUNS FOR DBA-LOCA FOR LONG-TERM NPSH (K=147, SWT=900F)..................................................

I-1 APPENDIX J: PLOTS FROM SHEX RUNS FOR DBA-LOCA FOR LONG-TERM NPSH (K=147, SWT=94-F)..................................................

J-1 APPENDIX K: PLOTS FROM SHEX RUNS FOR DBA-LOCA WITH DIRECT POOL COOLING (K=147, SWT=900F)................................................. K-1 APPENDIX L: PLOTS FROM SHEX RUNS FOR DBA-LOCA WITH DIRECT POOL COOLING (K=147, SWT=94-F).

................................................. L-1 iii

GE-NE-0000-0002-8817-01-R2 LIST OF TABLES Table Title Page 3-1 Key SHEX Input Values for DBA-LOCA with Direct Pool Cooling................ 3-15 3-2 Key SHEX Input Values for DBA-LOCA for Short-term NPSH.............

.......... 3-16 3-3 Key SHEX Input Values for Long-term DBA-LOCA with Containment Spray Cooling for NPSH..............................................

3-17 3-4 Key SHEX Input Values for Isolation Event, 0.01 ft2 and 0.1ft2 Break with Direct Pool Cooling.................................................

3-18 4-1 SAFER/GESTR Analysis Results.................................................

4-19 4-2 Peak Suppression Pool Temperature for Various Events with Direct Pool Cooling - RHR Heat Exchanger K=143.1 Btulsec-0F and 900F Service Water.4-20 (Additional tables are included in Appendices C through E.)

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GE-NE-000}-0002-8817-01-R2 LIST OF FIGURES Figure Title Page (Included in Appendices F through L.)

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GE-NE 000-0002-8817-01-R2 ACRONYMS AND ABBREVIATIONS ADS Automatic Depressurization System ANS American Nuclear Society BWR Boiling Water Reactor CRD Control Rod Drive CS Core Spray CST Condensate Storage Tank DBA Design Basis Accident ECCS Emergency Core Cooling System EOP Emergency Operating Procedure GE GE Nuclear Energy HEM Homogeneous Equilibrium Model HPCI High Pressure Coolant Injection K

RHR heat exchanger K-value LOCA Loss of Coolant Accident LPCI Low Pressure Coolant Injection LWL Low Water Level MSIV Main Steam Isolation Valve NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission RHR Residual Heat Removal RPV Reactor Pressure Vessel SRV Safety-Relief Valve SWT Service Water Temperature TAF Top of Active Fuel TS Technical Specifications UFSAR Updated Final Safety Analysis Report V

GE-NE-0000-0002-8817-01-R2

1. PROJECT OVERVIEW AND OBJECTIVES A containment response analysis was performed in support of the power rerate project and the analysis results were documented in References 1 and 2.

The power rerate containment analysis was performed for a reactor thermal power of 1880 MWt. The long-term analysis used ANS 5.1 nominal decay values, assuming a service water temperature of 90'F and a residual heat removal (RHR) heat exchanger K-value of 143.1 Btu/sec-0 F. All applicable containment requirements associated with the responses analyzed were met at power rerate conditions.

Reference 2 supplemented the results of the Reference 1 analysis, and provided DBA-LOCA (design basis accident - loss-of-coolant accident) evaluations at power rerate conditions for use in analyses of net positive suction head (NPSH) for the residual heat removal (RHR) and core spray pumps. The Reference 3 containment analysis used a containment cooling initiation time based on the maximum time that it takes to establish containment cooling following a loss-of-coolant accident or reactor isolation event, as determined by a SAFER/GESTR analysis. The premise in the Reference 3 analysis was that adequate core cooling must be established before the low-pressure coolant injection (LPCI)/RHR system can be placed into the containment-cooling mode. In Reference 3, a reactor isolation event with the high pressure coolant injection (HPCI) unavailable was determined to be the most limiting event with respect to the maximum time to establish containment cooling. The maximum time required to initiate containment cooling for the isolation event was calculated to be 48.54 minutes from the onset of the event, based on the SAFER/GESTR analysis of the event. This time consists of the time to reach the LPCI injection valve permissive pressure, plus the 5-minute LPCI interlock, plus the 400 seconds to complete system configuration (line-up).

The DBA-LOCA containment response was evaluated, using the 48.6-minute delay in containment cooling, as compared with a 10-minute delay assumed for the Reference 1 power rerate analysis. The purpose of this containment analysis was to determine the impact of the 48.54-minute delay in containment cooling on peak suppression pool 1-1

GE-NE0000-0002-8817-01-R2 temperature. It should be noted that the analysis is considered very conservative because the DBA-LOCA event was analyzed, using the maximum time delay in containment cooling, which was obtained for the isolation event.

The DBA-LOCA analysis with a 48.54-minute containment cooling initiation time resulted in peak suppression pool temperature above 1950F (the current design temperature for piping system attached to the torus), when the service water (SW) temperature was assumed to be 90TF. With 85F SW temperature, however, the same case resulted in peak suppression temperature below 195TF. Consequently, a reduction in the maximum allowable SW temperature on the ultimate heat sink (UHS) from 90T to 85TF was imposed by Monticello.

This project was initiated to restore the maximum acceptable UHS temperature to 900F (Reference 10). The project is also intended to address the following issues (Reference 10):

a. Delayed cooling time for other small breaks - The reactor isolation event can be considered to conservatively represent small breaks, as far as the maximum time to establish containment cooling is concerned. However, it will be necessary to confirm such is the case by analyzing additional small breaks.

b. Revised decay heat requirements - The containment analyses of References 1, 2 and 3 were calculated at 102% of 1880 MWt using a decay heat curve based on a the nominal ANSI/ANS 5.1-1979 decay heat with no uncertainty adders.

However, the NRC currently requires a 2-sigma adder to the ANS 5.1-1979 decay heat in containment analyses.

As described in Reference 1, the decay power time-history used for the Reference 1 analysis, using the nominal ANS 5.1-1979 decay heat at 102% of 1880 MWt, is roughly equivalent to the power corresponding to ANS 5.1-1979 with a 2 sigma adder at 102% of 1775 MWt. This provided justification at the time for using the results of the Reference 1 analysis to support a power rerate to 1775 MWt. However, it was understood that future analyses will include generation of a Monticello plant specific ANS 5.1-1979 + 2 sigma decay heat curve.

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GE-NE-0000-0002-8817-01-R2

c. GE SIL 636 - In June 2001, GE issued Service Information Letter (SIL) 636 (Reference 8) which describes a potential non-conservatism in decay heat calculations based on the ANSI/ANS-5.1-1979 standard. The non-conservatism results from failure to account for the cumulative effects of actinides other than 239U and 23Np, as well as activation products in structural materials. This non-conservatism can potentially impact the results of the Reference 1, 2 and 3 analyses.

Consequently, Nuclear Management Company (NMC) requested GE, by Reference 10, to perform a containment analysis based on updated plant data, addressing the above concerns. This report provides the results of the containment analysis that has been performed for Monticello Nuclear Generating Plant, per Reference 10.

Specifically, the objectives of this project are:

1. Update the long-term containment analysis input basis by using updated plant data, and decay heat values that are calculated with additional terms, consistent with SIL 636, and also by adding 2-sigma adders to nominal values for the decay heat values to be used for the analysis.

2. Perform a SAFERIGESTR analysis to determine the ADS activation and containment cooling initiation times for an reactor isolation event, 0.01 and 0.1 ft2 liquid line breaks at 102% of 1775 MWt core power. The power level and decay heat values used in this analysis will be consistent with those for the containment analysis.

3. Perform a long-term containment analysis for the isolation event and small breaks, using the ADS activation and containment cooling initiation times obtained from the SAFER/GESTR analysis.

4. Perform a long-term containment analysis for the DBA-LOCA with direct suppression pool cooling, using input values that maximize the containment pressure response.

5. Perform a short-term (<600 seconds) and long-term (beyond 600 seconds up to 12 days) containment analysis for the DBA-LOCA with containment spray cooling, 1-3

GE-NE-0000-0002-8817-01-R2 using input values minimizing the containment pressure response.

The analysis results will be provided to the NMC for input to the evaluation of NPSH for pumps taking suction from the suppression pool.

6. Determine which event is the limiting event, among the DBA-LOCA, small-break LOCAs and the isolation event, with respect to peak suppression pool temperature.

For the limiting event, perform a long-term containment analysis with an updated K-value of 147 Btu/sec-0F for the RHR heat exchanger (Reference 5), as compared with the K-value of 143.1 Btu/sec-'F used in the Reference 1 power rerate analysis. This analysis will be performed with two sets of input assumptions: a) direct pool cooling with input values maximizing the pressure response, and b) containment spray cooling with input values minimizing the pressure response for input to NPSH evaluation.

7. Determine the maximum acceptable service water temperature that keeps the peak suppression pool temperature below 196.7°F, which has been identified, by the NMC, as an acceptable piping temperature limit (Reference 13). For this purpose, a long-term containment analysis will be performed for the limiting event with the RHR heat exchanger K-value of 147 Btu/sec-0F. This analysis will be performed with two sets of input assumptions: a) direct pool cooling with input values maximizing the pressure response, and b) containment spray cooling with input values minimizing the pressure response for input to NPSH evaluation.

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2.

SUMMARY

AND CONCLUSIONS 2.1 Summary The following summarizes the results of the analyses performed for this project:

1. The long-term containment analysis input basis has been updated, as documented in Reference 6. The key input parameters are listed in Section 3 of this report. Appendix A provides the total core heat values (fission power

+ decay heat (consistent with SIL 636) + fuel relaxation energy + metal-water reaction energy) that were used in the containment analysis.

2. A SAFER/GESTR analysis has been performed to determine the ADS activation and containment cooling initiation times for the reactor isolation event, 0.01 and 0.1 ft2 liquid line breaks at 102% of 1775 MWt core power.

The containment cooling initiation times for the isolation event, 0.01 and 0.1 ft2 breaks were determined to be 45.81, 26.0 and 16.76 minutes, respectively.

Note that the containment cooling initiation time was determined to be 48.54 minutes in the Reference 3 SAFER/GESTR analysis, which was performed at 1775 MWt with the decay heat values obtained before SEL 636. The use of the 2%-higher initial power level and SIL-636 consistent decay values resulted in a reduction of the delay in containment cooling. For consistency between the SAFER analysis and containment analysis, the 45.81-minute value was used as input to the current containment analysis of the isolation event, which was performed at 102% of 1775 MWt core power.

3. A long-term containment analysis for the isolation event and small breaks with direct suppression pool coo1ing was performed with a service water temperature of 900F and the RHR K-value of 143.1 Btu/sec-0F. For all cases, the peak suppression pool temperature was below 195TF, as shown below.

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GE-NE-0000-0002-8817-01-R2 Reactor Isolation 45.81 194.0 One RHR loop with HPCI unavailable.

Reactor 45.81 167.0 Two RHR loops with HPCI Isolation unavailable.

0.01 26.0 190.0 One RHR loop with HPCI unavailable.

0.1 16.76 191.2 One RHR loop with HPCI

______unavailable.

Note: Reactor Isolation event with one RHR loop and inoperable HPCI bounds small break cases with HPCI available, as discussed in section 4.2.4.

4. A long-term containment analysis of the DBA-LOCA with direct suppression pool cooling, a 10-minute containment cooling initiation time, a service water temperature of 900F and the Reference 1 RHR K-value of 143.1 Btu/sec-0F, resulted in a peak suppression pool temperature of 195.60F.

5. A peak suppression pool temperature of 195.50F was obtained from a long-term containment analysis for the DBA-LOCA with containment spray cooling for input to the NPSH evaluation. This case was analyzed, assuming a 10-minute containment cooling initiation, with 90TF service water temperature and the Reference 1 RHR K-value of 143.1 Btu/sec-0F

6. The limiting event with respect to peak suppression pool temperature, among the DBA-LOCA, isolation event and small-break LOCAs, was determined to be the DBA-LOCA, as summarized above. When the updated RHR heat exchanger K-value of 147 Btu/sec-0F (Reference 5) was used with a service water temperature of 90TF, the peak suppression pool temperature for the limiting DBA-LOCA event was 194.10F with direct suppression pool cooling, and 194.20F with containment spray cooling for NPSH.

7. With the updated RHR heat exchanger K-value of 147 Btu/sec-0F and a service water temperature of 94TF, the peak suppression pool temperature for 2-2

GE-NE0000-0002-8817-01-R2 the limiting DBA-LOCA event was 196.50F with direct suppression pool cooling, and 196.20F with containment spray cooling for NPSH. Thus, the maximum acceptable service water temperature that would keep peak suppression pool temperature below the Reference 13 acceptable piping temperature limit of 196.70F was determined to be 940F with the updated RHR heat exchanger K-value of 147 Btulsec-0F.

2.2 Condusions Based upon the analysis performed with the updated plant data, and decay heat values (with 2-sigma adders) obtained with the method consistent with SIL 636, the following conclusions can be drawn for Monticello Nuclear Generating Plant

For the limiting DBA-LOCA event, the peak suppression pool temperature at 102% of 1775 MWt reactor power was above 195TF, when the Reference RHR heat exchanger K-value of 143.1 Btu/sec-0F was used with a service water temperature of 901F. With the updated RHR heat exchanger K-value of 147 Btu/sec-0F, however, the peak suppression pool temperature for the DBA-LOCA was below 195TF with the same service water temperature of 900F.

At 102% of 1775 MWt reactor power, the maximum acceptable service water temperature that would keep the peak suppression pool temperature below 196.70F for the limiting DBA-LOCA, is 940F with the updated RHR heat exchanger K-value of 147 Btu/sec-0F.

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3. EVALUATION 3.1 Method of Evaluation Long-term containment analyses are performed using the GE computer code SHEX, as was used for the Reference I power rerate evaluation, in compliance with the Reference 7 requirements. The key models used in the SHEX code are described in References 9 and 14.

The SHEX code was already used for the Monticello UFSAR analysis, which was preceded by confirmatory calculations (Reference 1) between the SHEX code and the NRC-approved HXSIZ code. Therefore, the use of the SHEX code for Monticello is within the current licensing analysis basis and complies with the NRC requirements.

Nevertheless, a benchmarking analysis has been performed by re-analyzing the DBA-LOCA analyzed in Reference 1, using the same decay heat values at the same initial core power, but with updated plant data.

Thus, the benchmarking analysis would show the impact of the plant data update and modeling update on the containment response. The benchmarking analysis, which is presented in Appendix B, shows that the current analysis resulted in a 0.20F increase in peak suppression pool temperature (the key containment response parameter) for the same core power and decay heat values.

The SHEX code is also used to evaluate the containment response for a reactor isolation event, and small breaks.

However, for these events, the ADS activation and RHR suppression pool cooling initiation times were determined from the SAFER/GESTR results for the same events, since more detailed vessel models in the SAFER/GESTR 3-1

GE-NE-0000-0002-8817-01-R2 code evaluate the vessel response more accurately. The ADS activation and suppression pool cooling initiation times determined as such are used as input to the SHEX code.

In the SAFER/GESTR analysis, the automatic activation of the ADS is inhibited and the operator is assumed to depressurize the reactor vessel by using the ADS valves when the water level outside the core shroud decreases to the top of the active fuel. The time of ADS activation is obtained directly from the SAFER/GESTR output. The time to initiate containment cooling is determined by selecting a longer time from the results for the following two possible paths:

3-2

GE-NE-0000-0002-8817-01-R2 Thus, the ADS activation and containment cooling initiation times were determined from the SAFERIGESTR analysis, and these values were used as input to containment analyses.

An isolation event is the most limiting case for determining the maximum time to establish pool cooling. For the DBA-LOCA, the vessel is depressurized and the core reflooded within a few minutes. Containment cooling can typically be established within about 10 minutes, and the current analysis assumed a 10-minute containment initiation time for the DBA-LOCA. As the assumed break size decreases, it takes a longer time to depressurize the vessel and reflood the core. The maximum time occurs for a zero size break case (a reactor isolation event) where the only inventory loss from the vessel is done by boil-off through the SRVs. LOCAs for 0.10 ft2 and 0.01ft2 recirculation line breaks were also analyzed to further prove the isolation event is the most limiting with regard to containment cooling initiation time. It is noted that for small breaks less than 0.10 ft2, any increase in the time to reflood due to loop selection logic failure is accounted for by use of an adder based upon a bounding case of no LPCI flow (i.e., all LPCI flow is assumed lost in the broken loop).

Another thing to note is that the current SAFER/GESTR analysis was performed at 102%

of 1775 MWt core power with the Reference 4 decay heat values (nominal plus 2 sigma adders) obtained with the method consistent with SIL 636. The power level and decay heat values are the same as those used for the current containment analysis, whereas the Reference 3 SAFERIGESTR analysis was performed at 1775 MWt core power and decay heat values obtained before SIL 636. The containment cooling initiation time from the current SAFERIGESTR analysis performed as such is about 3-minute shorter, compared with the Reference 3 SAFERIGESTR analysis.

More significantly, the current containment analysis for the isolation event is more realistic than the Reference 3 containment analysis.

The Reference 3 containment evaluation of the isolation event was done by analyzing a recirculation suction line break (DBA-LOCA break size) with a containment cooling initiation time of 48.54 minutes obtained from a SAFER/GESTR analysis of the isolation event. However, the current 3-3

GE-NE-0000-0002-8817-01-R2 containment analysis models the reactor system response to the isolation event, including the break size (near-zero break size modeled), the ADS activation, and closure of ADS valves at 50 psig.

This modeling approach, as explained above, was used for both direct suppression pool cooling and containment spray cooling.

It is noted that this 34

GE-NEoo00-0002-8817-01-R2 modeling update has a negligible impact on peak pool temperature, but provides a more realistic suppression chamber airspace pressure and temperature prediction.

3.2 Inputs and Assumptions for SAFER/GESTR Analysis of Isolation Event and Small Breaks The inputs to the current SAFERIGESTR analysis are essentially the same as those used in the SAFERIGESTR analysis performed for Reference 3, except for the the reactor thermal power, decay heat values, and the input values affected by the power level, which were obtained from heat balance data for 102% of 1775 MWt. (Reference 11). The current SAFER analysis, as in the current containment analysis, used decay heat values (nominal plus 2-sigma adders) obtained with the method consistent with SIL 636. Also, the key input values were confirmed by checking with Monticello Power Rerate OPL-4 (Reference 12). It is noted that there are some differences between the OPL-4 and the OPL-4A developed for this project. The major differences between the two are:

I A.

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GE-NE-0000-0002-8817-01-R2 The following assumptions define the events that were analyzed with SAFER/GESTR.

1. Adequate core cooling is defined by restoring reactor water level inside the core shroud to two-thirds core height (elevation of the jet pump suction) with one core spray operating. This is the definition of adequate core cooling used in the Monticello Emergency Operating Procedures (EOPs).

2. The jet pump water level from the SAFER run will be used to determine the time when the water level reaches the jet pump suction. This is the region in the vessel monitored by the fuel zone level instrument.

3. The operators will initiate containment cooling once adequate core cooling has been established. This assumption will result in the longest delay time in establishing containment cooling.

This assumption is consistent with the expected scenario for a large break LOCA where the vessel will rapidly depressurize and empty through the break, and then is quickly refilled by the automatic initiation of the ECCS. For a slowly developing event, such as reactor isolation, it is conservative to assume that containment cooling is delayed until the vessel is depressurized and adequate core cooling is established. For this type of event the operator would have time to initiate containment cooling before ECCS initiation occurs.

Since the additional cooling time removes energy from the containment, it is also conservative for this type of event to assume no containment cooling until after adequate core cooling is established.

4. The operators will depressurize the vessel using the ADS valves in accordance with the emergency operating procedures (EOPs) when the water level outside the shroud reaches the top of the active fuel. The downcomer elevation corresponding to an indicated level of -126 inches will be used to define the "top of active fuel." This is the action level defined in the EOPs.

5. It takes 400 seconds to realign the RHR system from the LPCI mode to containment cooling and to start the RHR service water system (Reference 3).

6. The LPCI isolation valve control contains a logic that prevents the valves from being opened, unless the reactor vessel pressure falls below the LPCI injection valve pressure permissive. Once the vessel pressure drops below the LPCI injection valve pressure permissive an open signal is given to the LPCI 3-6

~

GE-NE-0000-0002-8817-01-R2 injection valve and the open signal is maintained for at least the next 5 minutes, assuring the valve stays open.

8. The single failure assumed for this evaluation is a battery failure that eliminates the division of ECCS that has the HPCI. The remaining systems are two LPCIRHR pumps and one core spray pump. One RHR pump must be shed when the service water pump is started to initiate containment cooling.

9. It is assumed that no high pressure inventory makeup systems (feedwater, HPCI, RCIC) are available and that the water level in the vessel cannot be maintained without depressurizing and using the low pressure ECCS. If the high pressure makeup systems were available, the water level in the vessel will be maintained above the top of the core, thus assuring adequate core cooling at all times.

The operators would then be free to establish containment cooling at any time. The impact of HPCI operation on peak suppression pool temperature is qualitatively evaluated in Section 4.2.4.

10. Consistent with the containment analysis, 102% of the current licensed power level of 1775 MWt will be used in the analysis, as well as the ANS 5.1 +2 sigma decay heat from Reference 4.

1. A full core of GEl 1 fuel will be assumed for this analysis. The core is predominantly GEl 1.

3-7

GE-NE-0000-0002-8817-01-R2 Although GEl 1 GESTR data is used, the revised decay heat is based upon GE14 (Reference 4). However, Reference 4, which provides the new decay heat data, concludes that there is negligible difference between the decay heats for GE1 1 and GE 14.

3.3 Inputs and Assumptions for SHEX Containment Analysis The OPL-4A (Reference 6) lists the containment analysis input parameters used for the current analysis.

This section (Section 3.3) lists and discusses key inputs and assumptions used in the SHEX analysis for each of the events analyzed, starting with the DBA-LOCA with direct suppression cooling.

3.3.1 DBA-LOCA with Direct Suppression Pool Cooling Table 3-1 gives the values of major SHEX input parameters used for the current containment analysis of DBA-LOCA with direct pool cooling.

Key inputs and assumptions for SHEX analyses of the DBA-LOCA with direct suppression pool cooling are discussed below.

1. The DBA-LOCA is defined as a double-ended recirculation suction line break. The effective break area is 4.095 ft2. The RHR inter-tie is assumed open during the break and is included in the break size calculation. There is also a concurrent loss of offsite power and only minimum diesel power is available. Reactor scrams at time zero.

2. The reactor is operating at 102% of rated thermal power of 1775 MWt (i.e.

1810.5 MWt) with an initial reactor pressure of 1040 psia.

3. The reactor core power includes fission energy, fuel relaxation energy, metal-water reaction energy and ANS 5.1 + 2a decay heat for fuel applicable up to GE14 (see Appendix A). The decay heat values used address the SIL 636 issue.

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4. Reactor blowdown flow rates are based on the Homogeneous Equilibrium Model (HEM) described in Reference 15. This is the critical flow model used in the analysis of References 1, 2 and 3.

5. A hot portion of the feedwater inventory is transferred to the vessel to maximize the suppression pool temperature response.

6. MSIV closure starts at 0.5 seconds after the initiation of the event and full closure is achieved at 3.0 seconds after closure is initiated.

7. For the first 10 minutes following the accident, two LPCI pumps and one CS pump inject into the vessel. After 10 minutes, one CS remains available for vessel injection.

8. The initial suppression pool water volume corresponds to the Low Water Level (LWL) to maximize the suppression pool temperature response.

9. The initial drywell and suppression chamber airspace pressure is assumed to be at the scram setpoint pressure of 2 psig.

The initial dywell relative humidity is assumed to be 20% with an initial drywell temperature of 1350F.

An initial temperature of 90'F is assumed for the suppression pool. The initial suppression chamber airspace temperature is also assumed to be 900F with 100% relative humidity. Under such containment initial conditions, the total non-condensable gas mass in the containment (the drywell and suppression chamber airspace combined) is 18200 bm. The assumption of higher containment pressure and lower drywell humidity results in a larger amount of non-condensable gas mass in the containment, which in turn will provide a higher containment pressure response.

10. Drywell fan coolers are inactive.

11. Control rod drive flow is zero.

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GE-NE-0000-0002-881 7-0O1-R2

13. All core spray and LPCI/RHR pumps have 100% of their motor horsepower rating converted to pump heat which is added either to the RPV liquid or suppression pool water. This assumption is used to maximize the suppression pool temperature response.

14. Passive heat sinks corresponding to the drywell metal shell (including the vent system), and torus metal shell are modeled.

15. Heat transfer from the primary containment to the reactor building is conservatively neglected.

16. Six suppression chamber-to-drywell vacuum breakers are assumed to be active.

17. CST water inventory is not available for vessel makeup.

18. There is only one RHR loop with one heat exchanger available for containment cooling, starting at 10 minutes.

19. At 10 minutes, one LPCIIRHR pump is turned off and the remaining LPCIIRHR pump is realigned in suppression pool cooling mode with activation of RHR heat exchanger cooling. The RHR pump operates at 4000 gpm with flow through the heat exchanger. One RHR SW pump is aligned with the RHR heat exchanger. This configuration is maintained throughout the remainder of the accident. The corresponding RHR heat exchanger K-value is 143.1 Btu/sec-0F, based on the Reference 1 analysis.

In addition to this K-value, an updated K-value of 147 Btu/sec-0F is used in the analysis.

20. The RHR service water temperature is at the Reference 1 maximum value of 900F to maximize the suppression pool temperature. The DBA-LOCA was also analyzed at a service water temperature of 940F with the updated K-value of 147 Btu/sec-0F. (A 94F service water temperature was determined to be the maximum acceptable service water that would keep peak suppression pool below 196.70F with the updated K-value of 147 Btu/sec-0F during the DBA-LOCA (see Section 4.5).)

Thus, the DBA-LOCA with direct suppression pool cooling was analyzed for three different combinations of service water temperature and RHR heat exchanger K-value, as shown below, and the results are presented in Section 4.

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GE-NE-000-0002-8817-01-R2 SW Temperature RHR heat exchanger K-value (OF)

(Btu/sec-0F) 90 143.1 90 147 94 147 3.3.2 DBA-LOCA with Containment Spray Cooling for NPSH The DBA-LOCA containment response for NPSH evaluations is analyzed for two time periods: short-term (0-600 seconds) before the operator takes actions (such as initiation of containment cooling and throttling of the pump flow), and long-term (600 seconds-12 days) after initiation of containment cooling. Table 3-2 gives the values of the major SHEX input parameters used for the short-term DBA-LOCA analysis for NPSH, whereas the input values for the long-term DBA-LOCA analysis for NPSH are given in Table 3-3.

The inputs and assumptions for this analysis are essentially the same as those for the DBA-LOCA with direct suppression pool cooling listed in the previous section, except that some input values are selected to minimize the containment pressure. Specifically, the following assumptions are different from those for the DBA-LOCA with direct pool cooling. (Note that the following assumptions related are applicable to both short-term (0-600 seconds) and long-term (600 seconds to 12 days) NPSH DBA-LOCA analyses, unless specified for the event duration applicable.)

1. The initial drywell and suppression chamber airspace pressure is assumed to be 14.26 psia.

The initial drywell relative humidity is assumed to be 100%

with an initial drywell temperature of 1350F. An initial temperature of 900F is assumed for the suppression pool. The initial suppression chamber airspace temperature is also assumed to be 90'F with 100% relative humidity. Under such containment initial conditions, the total non-condensable gas mass in the containment is 14224 bm versus 18200 bm assumed to maximize the pressure response (Section 3.3.1 Assumption 9). The assumption of smaller containment pressure and lower drywell humidity results in a smaller amount of non-condensable gas mass in the containment, which in turn will provide a lower containment pressure response. This is done to minimize the pressure response for use in NPSH analyses.

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GE-NE-000O-0002-8817-01-R2

2. For the long-term (600 seconds-12 days) NPSH DBA-LOCA analysis, at 10 minutes one LPCI/RHR pump (out of the two operable pumps) is turned off and the remaining pump is realigned in containment spray (DW and WW sprays) cooling mode with activation of RHR heat exchanger cooling. The RHR pump operates at 4000 gpm with flow through the heat exchanger. One RHR SW pump is aligned with the RHR heat exchanger. This configuration is maintained throughout the remainder of the accident.

3. For the long-term (600 seconds-12 days) NPSH DBA-LOCA analysis, 3800 gpm, out of the total RHR 4000 gpm flow, goes to the drywell spray and the remaining 200 gpm goes to the suppression chamber spray.

5. For the short-term (0-600 seconds) NPSH DBA-LOCA analysis, two LPCS pumps are operable.

However, for the long-term (600 seconds-12 days) analysis, one LPCS pump is assumed to be operable as vessel makeup.

The DBA-LOCA with containment spray cooling for long-term NPSH was analyzed for three different combinations of service water temperature and RHR heat exchanger K-value, as shown below and the results are presented in Section 4. (It is noted that neither the SW temperature nor RHR heat exchanger K-value has any impact on the short-term NPSH DBA-LOCA results.)

SW Temperature RHR heat exchanger K-value (F)

(Btu/sec-0F) 90 143.1 90 147 94 147 3.3.3 Reactor Isolation and Small Breaks with Direct Suppression Pool Cooling A SHEX analysis was performed for a reactor isolation event and small breaks with direct suppression pool cooling, using ADS activation and containment cooling initiation 3-12

GE-NE-OOOO-0002-8817-01-R2 times obtained from the SAFERIGESTR analysis.

Table 3-4 gives the values of the major SHEX input parameters used for the containment analysis of these events with direct pool cooling. The analysis inputs and assumptions for this containment analysis are essentially the same as those for the DBA-LOCA with direct suppression pool cooling listed in Section 3.3.1, except for those listed below.

1. The break sizes analyzed are liquid breaks of 0.1 ft2, 0.01 ft2, and 0.0001 ft.

The 0.0001 ft2break is used to represent the reactor isolation event. This small break size (equivalent to about 4 gpm leakage at 1040 psia operating pressure) is modeled to account for the effect of the drywell heat sink in the suppression pool temperature response evaluation.

2. ADS initiation occurs as determined from the SAFERIGESTR analysis.

3. Two SRVs are used for the ADS functions, and they are closed at or below 50 psig.

4. Containment cooling (direct suppression pool cooling) is initiated, as determined from the SAFER/GESTR analysis (Table 4-1). One LPCIIRHR pump per loop is turned off and the remaining LPCL/RHR pump is realigned in suppression pool cooling mode with activation of RHR heat exchanger cooling. The RHR pump operates at 4000 gpm with flow through the heat exchanger. One RHR SW pump is aligned with the RHR heat exchanger.

This configuration is maintained throughout the remainder of the accident.

Only the Reference I RHR heat exchanger K-value of 143.1 Btu/sec-0F was analyzed. As shown in Section 4, the peak suppression temperature for the isolation event and also for small breaks was below 1950F. Therefore, an additional containment analysis with the updated K-value of 147 Btu/sec-0F was not necessary.

5. The service water temperature is, 90'F. An additional containment analysis with 94IF service water temperature was not necessary.

6. One CS pump remains available for injection into the vessel.

This configuration is maintained throughout the remainder of the accident.

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GE-NE-0000-0002-8817-01-R2

7. Two ECCS/RHR configurations: one-loop and two-loop operation, were analyzed for the isolation event represented by O.OOOift break. For other two break sizes, only one-loop configuration was analyzed.

8. Mechanistic modeling of heat and mass transfer between the suppression pool and the suppression chamber airspace was used from time zero.

9. HPCI is not available. A qualitative evaluation is performed for the case where HPCI is available to ascertain that the isolation event without HPCI is more limiting with respect to peak pool temperature (see Section 4.2.4).

Thus, the following cases are analyzed:

Break Sizes (ft2)

HPCI available?

Remarks Isolation No One RHR loop.

Event Isolation No Two RHR loops.

Event 0.01 No One RHR loop 0.1 No One RHR loop All sizes Yes Qualitative evaluation 3-14

GE-NE-0000-0002-8817-01-R2 Table 3-1: Key SHEX Input Values for DBA-LOCA with Direct Pool Cooling Parameter Initial Core Thermal Power (102% of rated)

Initial Reactor Vessel Dome Pressure Initial Drywell Free (Airspace) Volume Suppression Chamber Airspace Volume Initial Suppression Pool Volume Initial Drywell Pressure Initial Drywell Temperature Initial Drywell Relative Humidity Initial Suppression Chamber Pressure Initial Suppression Chamber Airspace and Pool Temperature Initial Suppression Chamber Relative Humidity Long-term Core Spray Flow Rate RHR Flow Rate RHR/CC Heat Exchanger K-Value Units MWt Value 1810.5 psia 3ft~

3ft~

3ft psia pF psia OF 1040 134,200 108,250 68,000 16.7 135 20 16.7 90 100 gpm gpm Btu/sec-0 F 4370 at 0 psid 3020 at 130 psid 4000 143.1 (current) 147 (updated) 90 94 (sensitivity) 600 RHRICC Service Water Temperature Suppression Pool Cooling Start Time sec 3-15

GE-NE-0000-0002-8817-01-R2 Table 3-2: Key SHEX Input Values for DBA-LOCA for Short-term NPSH Parameter Initial Core Thermal Power (102% of rated)

Initial Reactor Vessel Dome Pressure Initial Drywell Free (Airspace) Volume Suppression Chamber Airspace Volume Initial Suppression Pool Volume Initial Drywell Pressure Initial Drywell Temperature Initial Drywell Relative Humidity Initial Suppression Chamber Pressure Initial Suppression Chamber Airspace and Pool Temperature Initial Suppression Chamber Relative Humidity Core Spray Flow Rate per loop (2 loops available)

Total LPCI Flow Rate from 2 loops (2 pumps per loop) (2 loops available)

Units MWt Psia f>3 ftI 3

3 ft3 Psia OF psia OF gpm Value 1810.5 1040 134,200 108,250 68,000 14.26 135 100 14.26 90 100 4370 at 0 psid 3020 at 130 psid 17400 gpm 3-16

GE-NE-0000-0002-8817-01-R2 Table 3-3: Key SHEX Input Values for Long-term DBA-LOCA with Containment Spray Cooling for NPSH Parameter Initial Core Thermal Power (102% of rated)

Initial Reactor Vessel Dome Pressure Initial Drywell Free (Airspace) Volume Suppression Chamber Airspace Volume Initial Suppression Pool Volume Initial Drywell Pressure Initial Drywell Temperature Initial Drywell Relative Humidity Initial Suppression Chamber Pressure Initial Suppression Chamber Airspace and Pool Temperature Initial Suppression Chamber Relative Humidity Long-term Core Spray Flow Rate Drywell Spray Flow Rate Suppression Chamber Spray Flow Rate RHR/CC Heat Exchanger K-Value Units MMt Value 1810.5 psia 3ft 3ft f3 psia pF psia OF 1040 134,200 108,250 68,000 14.26 135 100 14.26 90 100 gpm gpm gpm Btu/sec-0F 4370 at 0 psid 3020 at 130 psid 3800 200 143.1 (current) 147 (updated) 90 94 (sensitivity) 600 RHRICC Service Water Temperature Suppression Pool Cooling Start Time sec 3-17

GE-NE-0000-0002-8817-O1-R2 Table 3-4: Key SHEX Input Values for Isolation Event, 0.01 ft2 and 0.1ft2 Break with Direct Pool Cooling Parameter Initial Core Thermal Power (102% of rated)

Initial Reactor Vessel Dome Pressure Initial Drywell Free (Airspace) Volume Suppression Chamber Airspace Volume Initial Suppression Pool Volume Initial Drywell Pressure Initial Drywell Temperature Initial Drywell Relative Humidity Initial Suppression Chamber Pressure Initial Suppression Chamber Airspace and Pool Temperature Initial Suppression Chamber Relative Humidity Long-term Core Spray Flow Rate RHR Flow Rate RHR Heat Exchanger K-Value RHR Service Water Temperature Number of ADS Valves Activated ADS Activation Time Containment Cooling Initiation Time Units MW Value 1810.5 psia ft3 fs3 ft3 3

psia pF psia OF 1040 134,200 108,250 68,000 16.7 135 20 16.7 90 100 gpm gpm Btu/sec-0F OF 4370 at 0 psid, 3020 at 130 psid 4000 143.1 (current) 90 2

sec min 1704 (isolation); 513 (0.01ff2);

58 (0.1 Wf) 45.81 (isolation);26 (0.01 ft2);

16.76 (0.1 f 2) 3-18

GE-NE0000-0002-8817-01-R2

4. EVALUATION RESULTS 4.1 ECCS-LOCA Results for Isolation Event and Small Breaks without HPCI The isolation event, and 0.01 and 0.1 ft2 breaks were analyzed, using SAFER, and the results are summarized in Table 4-1. The tabulated values include: a) ADS activation time, and b) two containment cooling initiation times. As discussed in Section 3.1, these values were determined from the results of a SAFER/GESTR analysis. Plots from the SAFER runs are included in Appendix F of this report. The SAFER results are discussed below.

Isolation Event The results for the isolation event are presented in Figures F.1-1 through F.1-6. Main steam line isolation and loss of off-site power were assumed to occur at the start of the event. The vessel pressure (Figure F.1-4) rapidly rises and is maintained between the safety relief valve opening and closing setpoints. The vessel water levels both outside and inside the shroud (Figures F.1-3 and F.1-2) decrease due to inventory loss from the boil-off from decay heat. When the vessel water level outside the shroud reaches the elevation of the top of the active fuel at 1704 seconds into the event, the operator initiates the Automatic Depressurization System. Flow through these valves continues until the valves close on low vessel pressure at 2344 seconds. ADS actuation causes the vessel pressure to rapidly decrease. The pressure permissives for the LPCS and LPCI are both reached at 2049 seconds. Once the vessel pressure drops below the shutoff head of the low pressure ECCS, reflooding flow begins to enter the vessel (Figure F.1-6). The LPCS begins injection at 2069 seconds and the LPCI begins injection at 2084 seconds. The vessel water level rapidly recovers due to the ECCS injection and the level to the jet pump suction is restored at 2170 seconds.

Therefore based on the jet pump level recovery, containment cooling would be established at 2170 + 400 = 2570 seconds (42.83 minutes). While based on the 5-minute LPCI interlock, containment cooling would be established at 2048.58 + 300 + 400 = 2748.58 seconds (45.81 minutes). These results are summarized in Table 4-1. Thus for this case the time to establish containment cooling is 4-1

GE-NE-0000-0002-8817-01-R2 limited by the LPCI interlock. This table also shows the previous results for the isolation event (Reference 3).

Compared with the Reference 3 analysis, a 2% higher power and ANS 5.1 + 2 sigma decay heat values (consistent with SIL 636) were used in the current analysis, as discussed in Section 3.1.

With these new input values, the containment cooling initiation time was determined to be 45.81 minutes, based on clearing of the 5-minute LPCI interlock. This time is about 3 minutes shorter than the Reference 3 analysis value of 48.54 minutes due to the higher power and new updated decay heat.

Recirculation Line Small Break Cases Additional SAFER runs were performed assuming recirculation line breaks of 0.01 ft2 and 0.10 ft2.

The system response to a small break is similar to the isolation case discussed above with the transient events occurring earlier due to the increased inventory mass loss through the break.

The results of the 0.01 ft2 break analysis are presented in Figures F.2-1 through F.2-6.

Main steam line isolation and loss of offsite power were assumed to occur at the start of the event. The vessel pressure (Figure F.2-4) rapidly rises and is maintained between the safety relief valve opening and closing setpoints. The vessel water levels both outside and inside the shroud (Figures F.2-3 and F.2-2) decrease due to inventory loss through the break and the boil-off from decay heat. When the vessel water level outside the shroud reaches the elevation of the top of the active fuel at 513 seconds into the event, the operator initiates the Automatic Depressurization System. Flow through these valves continues until they close on low vessel pressure at 1189 seconds. ADS actuation causes the vessel pressure to rapidly decrease. The pressure permissives for the LPCS and LPCI are both reached at 860 seconds. Once the vessel pressure drops below the shutoff head of the low pressure ECCS, reflooding flow begins to enter the vessel (Figure F.2-6). The LPCS begins injection at 880 seconds and the LPCI begins injection at 895 seconds. The vessel water level rapidly recovers due to the ECCS injection and the level to the jet pump suction is restored at 985 seconds. Therefore based on the jet pump level recovery, 4-2

GE-NE-0000-0002-8817-01-R2 containment cooling would be established at 985 + 400 = 1385 seconds (23.08 minutes).

While based on the 5 minute LPCI interlock, containment cooling would be established at 860.13 + 300 + 400 = 1560.13 seconds (26.00 minutes). It is noted that the 985-second time is approximate. This is because, for recirculation line breaks less than 0.10 ft2, loop selection logic failure (i.e., LPCI injection into the broken loop) should be considered for accurate PCT results. Since PCTs are not an output of this study, and since containment cooling initiation times based upon jet pump level recovery are significantly bounded by (a) LPCI pressure permissive plus 5 minute calculations and (b) smaller line break sizes approaching the isolation case, this additional analysis with subsequent SAFER overlays was deemed not necessary for this application. However, a bounding case of no LPCI flow (i.e., all LPCI flow assumed lost in the broken loop) resulted in a jet pump level recovery delay of 38 seconds. Adding this margin to the first result above yields establishment of containment cooling at 1423 seconds (23.72 minutes).

The results of the 0.1 ft2 break analysis are presented in Figures F.3-1 through F.3-6.

Main steam line isolation and loss of offsite power were assumed to occur at the start of the event. The vessel pressure (Figure F.3-4) rapidly rises and is maintained between the safety relief valve opening and closing setpoints. The vessel water levels both outside and inside the shroud (Figures F.3-3 and F.3-2) decrease due to inventory loss through the break and the boil-off from decay heat. When the vessel water level outside the shroud reaches the elevation of the top of the active fuel at 58 seconds into the event, the operator initiates the Automatic Depressurization System. Flow through these valves continues until they close on low vessel pressure at 530 seconds. ADS actuation causes the vessel pressure to rapidly decrease. The pressure permissives for the LPCS and LPCI are both reached at 305 seconds. Once the vessel pressure drops below the shutoff head of the low pressure ECCS, reflooding flow begins to enter the vessel (Figure F.3-6). The LPCS begins injection at 325 seconds and the LPCI begins injection at 340 seconds. The vessel water level rapidly recovers due to the ECCS injection and the level to the jet pump suction is restored at 459 seconds. Therefore based on the jet pump level recovery, containment cooling would be established at 459 + 400 = 859 seconds (14.32 minutes).

4-3

GE-NE-0000-0002-8817-01-R2 While based on the 5 minute LPCI interlock, containment cooling would be established at 305.49 + 300 + 400 = 1005.49 seconds (16.76 minutes).

The SAFER/GESTR results for the isolation event, and two LOCA cases are summarized in Table 4-1.

4.2 Containment Analysis Results for Isolation Event, Small Breaks and DBA-LOCA with 143.1 Btu/sec-0F K-Value and 900F Service Water A long-term containment analysis was performed for a reactor isolation event, and DBA-LOCA, using the SEX code, with the pre-updated RHR heat exchanger K-value of 143.1 Btu/sec-0F and 90'F service water temperature.

Essentially, the long-term containment responses to the DBA-LOCA and isolation events previously analyzed are re-analyzed in this section with the updated plant data, addressing concerns about the decay heat requirement (nominal plus 2-sigma uncertainty), and SIL 636. Additionally, two small-break LOCAs were analyzed. (See Section 1 for more detailed information on the decay heat requirement and SIL 636.)

The analysis was performed at 102% of 1775 MWt (current rated thermal power), using a decay heat profile based on the ANSI/ANS-5.1-1979 standard with an added conservatism corresponding to a two-sigma uncertainty, satisfying the revised decay heat requirement. The Reference 4 decay heat values based on the method consistent with SIL 636 were used. Two small-break (0.01 and 0.1ft2) LOCAs were also analyzed to ascertain that the isolation event is the limiting event (with respect to peak suppression pool temperature), among small-break LOCAs. The results presented in this section show that the DBA-LOCA is the limiting event with respect to peak suppression pool temperature, as opposed to the Reference 3 result that the isolation event was the limiting event.

4-4

GE-NE4000-0002-8817-01-R2 4.2.1 Isolation Event, Small-break LOCAs and DBA-LOCA with One RHR Loop A reactor isolation event and two small-break (0.01 and 0.1ft2 ) LOCAs were assumed to occur with a concurrent loss of offsite power.

As the MSIVs are closed, the RPV pressure is maintained between SRV closing and opening setpoints.

The vessel liquid inventory is depleted due to steam discharge through SRVs and break flow. As the vessel water level reaches the top of active fuel (TAF), the operator is assumed to initiate the ADS, using two SRVs. Low-pressure systems (one core spray pump and two LPCI pumps) are activated for vessel inventory makeup.

After ensuring that long-term core cooling will be maintained, the operator initiates the containment cooling (one RHR loop) by switching one LPCI pump to RHR pool cooling mode, turning off the other LPCI pump, and turning on one RHR service water pump.

A SAFER/GESTR analysis was performed for a reactor isolation event, 0.01 and 0.1ft2 breaks with one ECCS division available and without HPCI to determine the times of ADS activation and containment cooling initiation (see Table 4-1). Using the SHEX code, a long-term containment analysis was performed with the ADS activation and containment cooling initiation times determined from the SAFER/GESTR analysis, assuming direct suppression cooling (as opposed to containment spray cooling).

It is noted that the isolation event was analyzed by assuming that there is a 0.0001 ft2 liquid line break, which is equivalent to a leakage of approximately 4 gpm at operating pressure of 1040 psia. This small break was assumed to take into account the effect of the drywell shell (heat sink) on peak suppression pool temperature. In the SHEX calculation, the sensible heat of the vessel wall is transferred to the fluid inside the vessel and then to the suppression pool via SRV flow (and break flow if there is a break), assuming that heat transfer from the vessel surface to the drywell is negligible, relative to the energy transfer via break flow and SRV flow. Since the heat transfer from the vessel surface may not be negligible in determining the drywell temperature and pressure for the isolation event, the containment pressure response given in Figure G.1-2 was underestimated, and therefore, should not be used for design purposes.

However, the vessel wall sensible heat is properly accounted for in the calculation of suppression pool temperature response. It is 4-5

GE-NE-0000-0002-8817-01-R2 noted that the primary objective of this analysis is to evaluate the suppression pool temperature response.

The DBA-LOCA with direct suppression pool cooling was also analyzed for the same containment initial conditions.

For the DBA-LOCA, a double-ended break of a recirculation suction line is assumed to occur with a concurrent loss of off-site power.

Due to a diesel generator failure (by itself or caused by a battery failure, as a single failure), only one core spray pump and two LPCI pumps are available for vessel inventory makeup. At 600 seconds into the event, the operator initiates the containment cooling (one RHR loop) by switching one LPCI pump to RHR pool cooling mode, turning off the other LPCI pump, and turning on one RHR service water pump. Using the SHEX code, a long-term containment analysis was performed, assuming direct suppression cooling.

The peak values of suppression pool temperature for the isolation event, and 0.01 and 0.1 ft2 breaks and DBA-LOCA are given in Table 4-2. Plots from the SHEX runs for these events are included in Appendix G of this report.

As shown in this table, the highest peak suppression pool temperature of 195.60F was obtained for the DBA-LOCA among the five events analyzed. Thus, the DBA-LOCA is the limiting event with respect to peak suppression pool temperature, as compared to the Reference 3 analysis that showed the isolation event is more limiting. The following provides an explanation why such is the case.

First, the event sequences for the isolation event and DBA-LOCA were compared. As Figure G.1-3 shows, the RPV pressure for the isolation event oscillates between SRV opening and closing setpoints until around 1700 seconds, at which time the operator opens ADS valves.

Upon opening of the ADS valves, the RPV pressure decreases rapidly, and the suppression pool temperature increases rapidly due to steam discharge from the ADS valves (Figure G.l-l). The ADS valves are closed at the RPV pressure of 50 psig, and the RPV pressure is maintained around 50 psig for a long time after that.

Additionally, for the isolation event the vessel water level is maintained above the TAF by core spray flow. The suppression pool temperature continues to increase until around 4-6

GE-NE-0000-0002-8817-01-R2 32,000 seconds, at which time heat removal by the RHR is equal to heat addition to the suppression pool. The peak suppression pool temperature for this isolation event was 194.00F (see Table 4-2), and the RPV temperature at the time of peak suppression pool temperature was around 3001F, corresponding to the saturation temperature at 50 psig pressure.

As noted in Section 3.1, in Reference 3 the isolation event was analyzed in a very conservative manner, and the peak suppression pool temperature for the isolation event was obtained by analyzing a DBA-LOCA break size with the containment cooling initiation delayed from 10 minutes to 48.54 minutes.

The sequence of events for the two small breaks (0.01 and 0.1 ft2) is similar to that for the isolation event, except for the ADS and containment cooling initiation times (see Figures G.2-3 and G.3-3). The containment cooling was initiated earlier than the isolation event, and the peak suppression pool temperature for these events was lower than the peak value for the isolation event (see Table 4-2).

Additionally, the long-term containment response to the limiting DBA-LOCA with containment spray cooling was analyzed, using containment initial conditions and assumptions that minimize the containment pressure response, for input to NPSH evaluations.

The peak suppression pool temperature for the DBA-LOCA with such assumptions was 195.50F, as compared with 195.60F obtained for the direct suppression pool cooling case discussed above. The 0.1F difference between the two cases is due to the difference in cooling mode (direct pool cooling vs. spray cooling) and the modeling differences mentioned in Section 3.

4-7

GE-NE-0000-0002-881 7-01-R2 4.2.2 DBA-LOCA as Limiting Event The analysis results presented in Section 4.2.1 show that the DBA-LOCA is the limiting event with respect to peak suppression pool temperature. It is also noted that the peak suppression pool temperature with containment spray cooling for NPSH was slightly lower than the value obtained with direct suppression pool cooling.

As part of this project, additional DBA-LOCA analyses were performed with the updated RHR heat exchanger K-value of 147 Btu/sec-0F and 90'F service water temperature (see Section 4.4), and with the updated RHR heat exchanger K-value of 147 Btu/sec-0F and 940F service water temperature (see Section 4.5).

Since the peak suppression pool temperature is different (although slightly) between spray cooling for NPSH and direct suppression pool cooling, it was necessary to ascertain that the peak suppression pool temperature for either DBA-LOCA case is acceptable.

4.2.3 Isolation Event with Two RHR Loops A long-term containment analysis for the isolation event with two ECCS loops (two RHR containment cooling loops) was also performed, using the SHEX code. Relative to the one-ECCS loop case, the ADS activation time will be the same, whereas the containment cooling initiation time will be shorter with two ECCS loops (one CS pump and two LPCI pumps for each loop). The current analysis conservatively assumed that these times with two ECCS loops are the same as those for the one-ECCS loop case. However, for the containment analysis two loops were assumed to be available for both the ECCS and RHR containment cooling system. It was assumed that each RHR loop has one RHR pump and a heat exchanger K-value of 143.1 Btu/sec-0F and a service water temperature of 90'F.

As Table 4-3 shows, the peak suppression pool temperature for this event with two ECCS/RHR loops was calculated to be 167.00F, as compared with 194.00F obtained with one ECCS/RHR loop; a 270F reduction in peak suppression pool temperature.

Plots from the SHEX run for this case are given in Figures G.5-1 through G.5-3.

4-8

GE-NE0000-0002-8817-01-R2 4.2.4 Containment Response to LOCAs with Operable HPCI In Section 4.2.1, the containment response to three events with HPCI unavailable (a reactor vessel isolation event, and two small breaks) was evaluated, using the ADS activation and pool cooling initiation times obtained from the SAFER/GESTR analysis.

The results show that the isolation event without HPCI resulted in highest peak suppression pool temperature among the three events. Since the peak value for the isolation event was bounded by the DBA-LOCA peak value, the containment response for small breaks without HPCI is bounded by the DBA-LOCA result, as far as the suppression pool temperature response is concerned.

To ascertain that the same conclusion can be drawn for small breaks with HPCI available, the containment response for such cases is evaluated qualitatively in this section.

When the HPCI system is in operation during a postulated LOCA, three different scenarios depending upon the LOCA break size can be conceived, as follows:

Small breaks within HPCI capacity Large breaks above HPCI capacity Intermediate breaks comparable to HPCI capacity The impact of HPCI operation on the suppression pool temperature response is assessed qualitatively for the above three scenarios, by comparing with the scenario for the isolation event without HPCI, as explained below.

a) Small break within HPCI capacity - For a long period of time, the vessel water level can be maintained above the top of active fuel (TAF) with HPCI flow and vessel energy including decay heat will be released into the primary containment. As a result, the suppression pool temperature will increase above 901F, an Emergency Operating Procedure (EOP) entry condition.

The suppression pool-cooling mode of RHR will be initiated in accordance with the EOPs and suppression pool cooling will be initiated within 10 minutes into the event. This case with HPCI system available and the break 4-9

GE-NE-0000-0002-8817-01-R2 size within HPCI flow capacity is clearly less limiting than the isolation event without HPCI, in which the RHR pool cooling is initiated at 45.81 minutes.

b) Large break above HPCI capacity - For this case, the break flow exceeds HPCI makeup, and the water level will drop to the TAF relatively quickly.

ADS will be initiated earlier than the isolation event. The vessel reflood will occur earlier compared with the isolation event without HPCI. Therefore, the delay in containment cooling will be less, and this case with HPCI will be no worse than the isolation event without LPCI from a pool heat up standpoint.

c) Intermediate breaks comparable to HPCI capacity - It's possible that certain break sizes will be large enough that HPCI will not be able to maintain the vessel water level above the TAF, but small enough that the time it takes for the level to drop to the TAF could be longer than the time for the isolation event without HPCI. In this case, the RHR suppression pool cooling will be initiated within 10 minutes in accordance with the EOPs, as the pool temperature will be above 90'F shortly after initiation of the event, due to the release of vessel energy and decay heat into the pool.

4-10

GE-NE-0000-0002-8817-01-R2 4.3 DBA-LOCA Analysis for Short-term NPSH Evaluation The DBA-LOCA containment response for NPSH evaluations is analyzed for two time periods: short-term (before 600 seconds) before containment cooling is initiated, and long-term (after 600 seconds) after initiation of containment cooling. For this analysis for input to NPSH evaluations, the containment initial conditions are selected such that the containment pressure response can be minimized, as pointed out in Section 3.3.2.

This section presents the short-term NPSH DBA-LOCA results. Since the short-term DBA-LOCA for NPSH is performed for the time period before initiation of containment cooling, the RHR heat exchanger K-value and service water temperature have no impact on the short-term NPSH DBA-LOCA analysis.

Table C-1 of Appendix C provides time versus suppression pool temperature, suppression chamber airspace pressure with and without leakage, and suppression pool volume.

The 4-11

GE-NE-0000-0002-88 17-01-R2 leakage rate is specified in Reference 6. It is noted that the tabulated values show no differences between the suppression chamber airspace pressures with and without leakage, since the 1.2/o/day leakage during the 600-second time period is negligible.

Appendix H provides plots from the SHEX run for this case.

4.4 DBA-LOCA with 147 Btu/sec- 0F K-Value and 901F Service Water The NMC performed an extensive analysis to update the RHR heat exchanger K-value with the use of updated data, and provided GE, by Reference 5, with a new updated value of 147 Btu/sec-0F for use in the containment analysis of the limiting DBA-LOCA. Using this updated K-value, a long-term containment analysis of the DBA-LOCA was performed for two cases: a) direct suppression pool cooling, b) containment spray cooling for input to NPSH evaluations. This section presents the results of the long-term DBA-LOCA containment analysis performed with the updated RHR heat exchanger K-value of 147 Btu/sec-0F and 90'F service water temperature.

The analysis was performed at 102% of 1775 MWt (current rated thermal power), using the Reference 4 decay heat profile based on the ANSIIANS-5.1-1979 standard with an added conservatism corresponding to a two-sigma uncertainty (see Appendix A for the core heat values used in the analysis).

It is noted that the Reference 4 decay heat values were obtained with the method consistent with SIL 636.

4.4.1 DBA-LOCA with Direct Suppression Pool Cooling This long-term DBA-LOCA analysis is performed to maximize the long-term pool temperature response, while maximizing the long-term pressure response. To maximize the pressure response, the maximum initial containment pressure is assumed, along with the minimum value for the initial drywell humidity.

Containment cooling in the suppression pool cooling mode was assumed to start at 600 seconds. Only one RHR loop was assumed to be available with the updated RHR heat exchanger K-value of 147 Btu/sec-0F.

The service water temperature for the RHR was 4-12

GE-NE-0000-0002-8817-01-R2 assumed to be 90'F.

The inputs and assumptions used for the SHEX analysis of this event are listed in Section 3.3.1.

The peak suppression pool temperature of 194.10F occurred at 36,135 seconds into the event. This peak value is below the existing power rerate design temperature of 1950F for the piping attached to the torus. (Note that in a recent evaluation (Reference 13) a piping temperature increase to 196.70F was found to be acceptable (see Section 4.5).)

The long-term (after 600 seconds) suppression chamber airspace pressure was 18.3 psig, which is well below the suppression design pressure of 56 psig (Reference 1). The containment pressure and temperature responses are plotted in Figures K-1 through K-4 of Appendix K.

4.4.2 DBA-LOCA with Containment Spray Cooling for NPSH The DBA-LOCA containment response for NPSH evaluations is analyzed for two time periods: short-term (before 600 seconds), and long-term (after 600 seconds). For this analysis for input to NPSH evaluations, the containment initial conditions are selected such that the containment pressure response can be minimized. The short-term NPSH DBA-LOCA analysis results, which are independent of RHR heat exchanger K-value and service water temperature, are presented in Section 4.3. This section presents the long-term NPSH DBA-LOCA analysis results.

For the long-term NPSH DBA-LOCA analysis, only one RHR loop was assumed to be available with the updated RHR heat exchanger K-value of 147 Btu/sec-0F. Containment cooling in the containment spray mode is initiated at 600 seconds with a service water temperature of 900F. The inputs and assumptions used for the SHEX analysis of this event are listed in Section 3.3.2. Table D-1 of Appendix D provides the suppression chamber airspace pressure, suppression pool temperature and suppression pool volume responses vs. time for a one-day time period. The results for a 12-day time period are given in Table D-2.

These tables provide two values for the suppression chamber airspace pressure: one with leakage effects considered and the other without leakage.

The suppression chamber airspace pressure was calculated without leakage, and the 4-13

GE-NE-0000-0002-88 17-01-R2 suppression chamber airspace pressure with leakage was calculated by subtracting non-condensable gas mass at the rate of 1.2% per day from the non-leakage result. The leakage rate is specified in Reference 6. It is noted that the difference between the suppression chamber airspace pressures with and without leakage at one day into the event is less than 0.2 psi, since the 1.20/o/day leakage during the one-day period is not significant.

On the other hand, the leakage effect on the suppression chamber airspace pressure at 12 days was more than 2 psi.

Figures 1.1-1 through 1.1-4 present the suppression pool temperature, suppression chamber airspace pressure with and without leakage, and suppression pool responses for the one-day period. The results for the 12-day period are plotted in Figures I.2-1 through 1.2-4.

The peak suppression pool temperature from this NPSH DBA-LOCA analysis was 194.21F, which is below the existing the piping design temperature of 1951F.

The 194.20F peak value is 0.1F higher than 194.1 F obtained for the direct suppression pool cooling case. The 0.1F difference between the two cases is due to the difference in cooling mode (direct pool cooling vs. spray cooling) and the modeling difference, as discussed in Section 4.2.1.

4.4.3 Impact of 60-second Delay in Realignment to Containment Cooling Mode The concept that LPCI injection through the heat exchanges is equivalent to placing RHR into suppression pool cooling was assessed for its impact during transfer from LPCI mode to suppression pool cooling mode of RHR. During the most limiting DBA-LOCA scenario, after adequate core cooling is achieved, an RHR pump is secured and RHRSW pump is placed into service. At this time the remaining RHR pump is providing LPCI flow through the heat exchanger. In order to complete the transfer from LPCI mode to suppression pooling cooling mode, the suppression pool return valves have to be opened and the LPCI injection valves have to be closed. Operating procedures require that the suppression pool return valves be opened then closing the LPCI injection valves. The LPCI injection valves closure times are approximately 60 seconds. The net effect is that equivalent long-term containment cooling has been implemented prior to closure of the LPCI injection valves.

4-14

GE-NE-0000-0002-8817-01-R2 In order to specifically evaluate this concept in a bounding way, an analysis was performed that assumed it would take 660 seconds for the operator to realign the LPCIIRHR system to the containment-cooling mode for the LPCI mode during the DBA-LOCA.

The impact of the 60 second time delay on the peak suppression pool temperature would be negligible (<0.1F).

Reference 1 evaluated an alternative mode available to Monticello to achieve long-term containment cooling. This alternative mode would be to keep the RHR pumps in LPCI injection mode. The RHR flow is routed through the heat exchangers, injected into the vessel via the unbroken recirculation loop, passes through the lower plenum, spills out through the break, and returns to the suppression pool. Since LPCI injection through the heat exchanger in this mode would not be assumed to begin until 10 minutes into the event, the vessel would, by this time, be depressurized to the containment pressure.

Therefore, the LPCI pumps flow rates to the vessel, and the flow through the heat exchanger, would be equal to the flow rate assumed with the RHR pumps in pool cooling mode. Since the flow through the heat exchanger would not be expected to change, the heat exchanger performance with the RHR pumps maintained in LPCI mode would be the same as that assumed with RHR pool cooling mode in the analysis. Since heat exchanger performance does not change with the RHR pumps in LPCI injection mode, the energy removed for primary containment is not expected to change, and the difference in the peak suppression pool temperature between operating the RHR pumps in LPCI mode versus suppression cooling mode is expected to change by less than one degree Fahrenheit (dTF).

Thus, even with 1PF increase the DBA-LOCA peak suppression temperature for this alternative mode is below the maximum acceptable piping temperature of 196.7F given in Reference 13.

4.5 DBA-LOCA with 147 Btulsec-°F K-Value and 941F Service Water The Reference 13 letter provides the results from an analysis that was performed to assess the impact on the piping design temperature for all lines communicating with the torus, due to an increase in the suppression pool temperature from 1950F to 196.70F. It 4-15

GE-NE-0000-0002-8817-01-R2 was concluded, in Reference 13, that all the affected lines, and their associated supports, penetrations and nozzles met Code acceptance criteria for the elevated suppression pool temperature of 196.70F. A sensitivity analysis of the long-term DBA-LOCA response has been performed to determine the maximum acceptable service water temperature that would keep peak suppression temperature below 196.70F with the updated RHR heat exchanger K-value of 147 Btu/sec-0F. The sensitivity analysis was performed for two cases: a) direct suppression pool cooling, and b) containment spray cooling for input to NPSH evaluations. For this sensitivity analysis, the initial suppression pool temperature was assumed to be 90'F, the Monticello Technical Specification limit (Reference 5).

Based on sensitivity analyses, the maximum acceptable service water temperature was determined to be 940F, and this section presents the results of the long-term DBA-LOCA containment analysis performed with the updated RHR heat exchanger K-value of 147 Btulsec-0F and 940F service water temperature. As shown later in this section, the peak suppression pool temperature was below 196.70F under such conditions, thus confirming that a service water temperature of 940F with the RHR heat exchanger K-value of 147 Btu/sec-0F is the maximum acceptable service temperature that would keep peak suppression pool temperature below 196.7'F for the limiting DBA-LOCA.

The analysis was performed at 102% of 1775 MWt (current rated thermal power), using a decay heat profile based on the ANSIIANS-5.1-1979 standard with an added conservatism corresponding to a two-sigma uncertainty. The Reference 4 decay heat values based on the method consistent with SIL 636 were used.

4.5.1 DBA-LOCA with Direct Suppression Pool Cooling This long-term DBA-LOCA analysis is performed to maximize the long-term pool temperature response, while maximizing the long-term pressure response. To maximize the pressure response, the maximum initial containment pressure is assumed, along with the minimum value for the initial drywell humidity.

4-16

GE-NE-0000-0002-8817-01-R2 Containment cooling in the suppression pool cooling mode was assumed to start at 600 seconds. Only one RHR loop was assumed to be available with the updated RHR heat exchanger K-value of 147 Btu/sec-0F. The service water temperature for the RHR was assumed to be 940F, as determined to be acceptable based on sensitivity analyses. The inputs and assumptions used for the SHEX analysis of this event are listed in Section 3.3.1.

The peak suppression pool temperature of 196.50F occurred at 37,104 seconds into the event. This peak value is below the acceptable temperature of 196.70F for the piping attached to the torus (Reference 13). The long-term (after 600 seconds) suppression chamber airspace pressure was 19 psig, which is well below the suppression design pressure of 56 psig (Reference 1). The containment pressure and temperature responses are plotted in Figures L-1 through L-4 of Appendix L.

4.5.2 DBA-LOCA with Containment Spray Cooling for NPSH The DBA-LOCA containment response for NPSH evaluations is analyzed for two time periods: short-term (before 600 seconds), and long-term (after 600 seconds). For this analysis for input to NPSH evaluations, the containment initial conditions are selected such that the containment pressure response can be minimized. The short-term NPSH DBA-LOCA analysis results, which are applicable to any RHR heat exchanger K-value and service water temperature, are presented in Section 4.3. This section presents the long-term NPSH DBA-LOCA analysis results.

For the long-term NPSH DBA-LOCA analysis, only one RHR loop was assumed to be available with the updated RHR heat exchanger K-value of 147 Btu/sec-0F. Containment cooling in the containment spray mode is initiated at 600 seconds with a service water of 941F. The inputs and assumptions used for the SHEX analysis of this event are listed in Section 3.3.2. Table E-1 of Appendix E provides the suppression chamber airspace pressure, suppression pool temperature and suppression pool volume responses vs. time for a one-day time period. The results for a 12-day time period are given in Table E-2.

These tables provide two values for the suppression chamber airspace pressure: one with 4-17

GE-NE-0000-0002-88 17-01-R2 leakage effects considered and the other without leakage. Figures J.1-1 through J.14 present the suppression pool temperature, suppression chamber airspace pressure with and without leakage, and suppression pool responses for the one-day period. The results for the 12-day period are plotted in Figures J.2-1 through J.24.

The peak suppression pool temperature from this NPSH DBA-LOCA analysis was 196.20F, which is below the acceptable piping temperature of 196.70F. The slightly lower peak suppression pool temperature with NPSH-related assumptions, relative to the 196.50F for the direct pool cooling case, is due to the difference in cooling mode (direct pool cooling vs. spray cooling) and the modeling difference, as discussed in Section 4.2.1.

4-18

GE-NE-0000-0002-8817-01-R2 Table 4-1: SAFER/GESTR Analysis Results Case Description Reactor Reactor 0 01 f 0.10 ft2 Isolation Isolation recirculation recirculation line break line break Power (MWt) 1775 1810.5 1810.5 1810.5 Decay Heat Original ANS 5.1+2 ANS 5.1 + 2 c ANS 5.1+ 2 a Single Failure Battery Battery Battery Battery Jet Pump Recovery.

2336 2170 985 459 Time (sec)

Cont. Cool. Init.

45.60 42.83 23.08 (loop 14.32 (Based on Jet Pump selection logic Recovery Time +400°)

failure would add sec.) (minutes) less than 0.64 LPCI Pressure 2212.62 2048.58 860.13 305.49 Permis. Time (sec)

Cont. Cool. Init.

48.54 45.81 26.00 16.76 (Based on LPCI Pressure Permissive + 700w) sec.)

(minutes)

ADS Initiation time 1867 1704 513 58 (sec.)

I I

I

1) It takes 400 seconds to realign the RHR system from the LPCI mode to containment cooling and to start the RHR service water system (Reference 3).

2) The RHR system will remain locked in the LPCI mode for five minutes (300 seconds) following the pressure permissive. Plus an additional 400 seconds for note 1.

4-19

GE-NE-0000-0002-8817-01-R2 Table 4-2: Peak Suppression Pool Temperature for Various Events with Direct Pool Cooling

- RHR Heat Exchanger K=143.1 Btu/sec-0F and 900F Service Water Break Sizes Peak Suppression (ft2)

Pool Temperature Remarks

~~~~(OF)

Reactor 194.0 One RHR loop with inoperable HPCI Isolation Reactor 167.0 Two RHR loops with inoperable HPCI Isolation 0.01 190.0 One RHR loop with inoperable HPCI 0.1 191.2 One RHR loop with inoperable HPCI DBA-LOCA 195.6 One RHR loop with inoperable HPCI 4-20

GE-NE-0ooo-0002-8817-01-R2

5. REFERENCES

1.

GE-NE-T2300721-00-01 "Containment Response Evaluation Task 6.0" April 2, 1999. (Monticello Power Rerate Task Report).

2.

GE-NE-T23-00731-2 "Monticello Nuclear Generating Plant LOCA Containment Analyses for Use in Evaluation of NPSH for the RHR and Core Spray Pumps,"

dated June 1997.

3.

Letter, NSA 01-134, S. Mintz to G. Maxwell, "Monticello Nuclear Power Station -

Response to NMC Question Regarding Maximum Time to Vessel Reflood and Initiation of Containment Cooling," dated March 25, 2001.

4.

Letter, NSA-01-293, C. L. Martin to G. E. Maxwell, "Decay Heat Tables for Monticello Nuclear Generating Plant," dated July 10, 2001.

5.

Letter, A. Wojchouski to K. Narayan, "Containment Analyses for Monticello Heat Exchanger K-Value and Suppression Pool Temperature," dated May 9, 2002.

6.

Monticello Final OPL-4A attached to Letter, NSA 02-241, S. Mintz to G. Maxwell, "Monticello Containment Analysis Project - Final OPL-4A," dated April 12, 2002.

(Per WIN KHN-008 dated June 14, 2002, the initial drywell temperature used in the analysis of the DBA-LOCA for input to NPSH was changed to 1350F from the 1500F value specified in the OPL-4A dated April 12, 2002.)

7.

"Use of SHEX Computer Program and ANSUIANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis," Letter from Ashok Thadani (NRC) to Gary L. Sozzi (GE), July 13, 1993.

8.

GE Nuclear Energy Services Information Letter (SIL) Number 636, Revision 1, June 6, 2001.

9.

"The General Electric Mark III Pressure Suppression Containment System Analytical Model," NEDO-20533, June 1974.

10.

NMC PO 439, GE Proposal 523-JX7EY-EK1, Monticello Containment Analyses Project, Project Work Plan, Revision 1, February 2002.

5-1

GE-NE-0000-0002-8817-01-R2

11.

Letter, GLN-95-020, PT Tran (GENE) to SJ Hammer (NSP), "Reactor Heat Balances for Monticello Power Rerate (Task 2.1)," dated July 12, 1995.

12.

Letter, SJ Hammer (NSP) to PT Tran (GENE), "Emergency Core Cooling Parameters for Use in Monticello SAFER/GESTR Power Rerate Analyses - Task 7.5 (Approved copy)," dated November 21, 1997.

13.

Letter, Joe Attwood (Automated Engineering Service Corp.) to A. Wojchouski (NMC), "Post-LOCA Torus Water Temperature Increase," dated August 15, 2001.

14.

"The GE Pressure Suppression Containment Analytical Model," NEDO-10320, April 1971.

15.

"Maximum Discharge of Liquid-Vapor Mixtures from Vessels," NEDO-21052, September 1975.

5-2

GE-NE-OOO0-0002-8817-01-R2 APPENDIX A: CORE HEAT VALUES USED IN CONTAINMENT ANALYSIS The total core heat used in the containment analysis consists of shutdown power (composed of fission power and decay heat), fuel relaxation energy and metal-water reaction energy.

The three components and the total core heat, which are normalized against 102% of 1775 MWt, are given in Table A-1.

Reference A-1 provides the shutdown power (decay heat +fission power) values. The decay heat was calculated, using the ANSI/ANS 5.1-1979 standard, with the method consistent with SIL 636 (Reference A-2). The decay heat component of the shutdown power used in the current analysis was calculated by adding a 2-sigma uncertainty to nominal decay heat.

The metal-water reaction occurs as a result of a large LOCA. A LOCA results in a rapid drop in the vessel water level, before any makeup water can be injected, exposing part of the fuel bundles and the external zirconium claddings. This exposure leads to higher temperatures, resulting in reaction of the claddings with the surrounding water. This metal-water reaction energy is applied to all of the current containment analysis cases including the isolation event.

Note that Regulatory Guide 1.7 specifies that the metal-water reaction energy shall be based on 0.00023 inches of cladding reacting with water or 5 times the maximum amount calculated per NRC approved ECCS evaluation model (currently, GE's SAFER code) for demonstrating compliance with 0CFR50.46, Paragraph (b)(3), whichever is greater.

A-1

GE-NE-0000-0002-8817-01-R2 References Used in Appendix A A-1.

Letter, NSA-01-293, C. L. Martin to G. E. Maxwell, "Decay Heat Tables for Monticello Nuclear Generating Plant," dated July 10, 2001.

A-2.

GE Nuclear Energy Services Information Letter (SIL) Number 636, Revision 1, June 6, 2001.

A-2

GE-NE-0000-0002-8817-01-R2 Table A-1: Normalized Core Heat Values Used in Containment Analysis Time Fuel Metal-water Total Core Heat (sec)

Relaxation Reaction I_

I__

_ _ _ _ _ _ _ _ _ _ _I__ _ _ _ _ _ _ _

_ _ _ _ _ _ _ _ _ _ _ _ __I_ _ _ _ _ _ _ _ _ _ _ _ _ _

4 4

4

.4

.t

.4 1

4 4

.4

.4 1-

.4 4

4

4.

4

.4

4.

4 4

.4 1

I.

.4

4.

4

4.

.4 1-

.4

4.

4 I<

I I=

I I I

I~~~~~~~~~~L Note: Normalized against 102% of 1775 MWt.

A-3

GE-NEX)OO-0002-8817-01-R2 APPENDIX B: SHEX BENCHMARKING ANALYSIS FOR DBA-LOCA The SHEX code was already used for the MNGP UFSAR analysis, which was preceded by confirmatory calculations (Reference B-i) between the SHEX code and the NRC-approved HXSIZ code. Therefore, the use of the SHEX code for MNGP is within the current licensing basis analysis and complies with the NRC requirements.

Relative to the values used in the Reference B-1 power rerate analysis, the input values for the long-term containment analysis (OPL-4A form) have been updated for the current analysis, as documented in Reference B-2. In addition, the modeling of heat transfer between the suppression chamber airspace and pool for the DBA-LOCA is updated for the current analysis, as described in Reference B-2.

Benchmark calculations have been performed to quantify the impact of the OPL-4A input and modeling update on the SHEX results by re-analyzing the DBA-LOCA case analyzed in Reference B-1. The re-analysis of the DBA-LOCA case for benchmarking assumes exactly the same core power and decay heat values as used in Reference B-i, but with the updated OPL-4A input values and modeling.

Table B-1 shows the difference in peak suppression pool temperature (the key containment response parameter for the current analysis) between the Reference B-1 analysis and the current benchmarking analysis. As this table shows, the benchmarking analysis results in 0.20F increase in peak suppression pool temperature. This comparison indicates that the OPL-4A input and modeling update results in a slightly more conservative prediction of peak suppression pool temperature.

References Used in Appendix B B-1

GE-NE-0000-0002-8817-01-R2 B-1.

GE-NE-T2300721-00-01 "Containment Response Evaluation Task 6.0" April 2, 1999. (Monticello Power Rerate Task Report).

B-2 MONTICELLO OPL-4A, attached to NSA 02-241, S. Mintz to G. Maxwell "Monticello Containment Analysis Project -Final OPL-4A," dated 4/12/02.

B-2

GE-NE-0000-0002-8817-01-R2 Table B-1: Benchmarking Results with Reference B-I Core Power and Decay Heat Values Peak suppression pool temperature (OF) 193.4 193.6 Time of peak suppression pool temperature (sec) 31,937 34,082 B-3

GE-NE-0000-0002-8817-01-R2 APPENDIX C TABLE FOR SHORT-TERM CONTAINMENT RESPONSE TO DBA-LOCA FOR INPUT TO NPSH Table Title Page C-l Short-Term (<600 seconds) Response C-2 C-1

GE-NE-0000-0002-8817-01-R2 Table C-1 Short-Term (<600 seconds) Response TIME (Seconds) 0.0 33.7 53.4 62.9 73.5 86.9 98.8 108.6 118.0 127.4 137.0 148.2 160.0 174.1 191.4 210.2 229.4 249.4 271.6 294.7 319.1 344.4 368.4 393.9 419.0 441.6 465.7 489.1 512.4 535.4 559.6 583.2 Wetwell Pressure with Leakage (PSIA) 14.26 32.70 32.66 32.70 32.79 32.93 32.94 32.90 32.22 30.29 27.71 24.83 22.38 20.32 19.03 18.23 17.72 17.38 17.12 16.95 16.80 16.63 16.51 16.42 16.36 16.31 16.32 16.33 16.35 16.37 16.39 16.42 Wetwell Pressure without Leakage (PSIA) 14.26 32.70 32.66 32.70 32.79 32.93 32.94 32.90 32.22 30.29 27.71 24.83 22.38 20.32 19.03 18.23 17.72 17.38 17.12 16.95 16.80 16.63 16.51 16.42 16.36 16.31 16.32 16.33 16.35 16.37 16.39 16.42 Suppression Pool Temperature (F) 90.0 132.2 134.6 134.9 135.8 137.2 138.5 139.1 139.4 139.9 140.5 141.3 142.0 142.9 143.8 144.7 145.3 146.0 146.6 147.1 147.7 148.3 148.7 149.1 149.5 149.8 150.1 150.3 150.6 150.8 151.1 151.3 Suppression Pool Volume (CU FT) 68000 74280 74430 74660 74980 75430 75550 75560 75560 75550 75550 75570 75580 75610 75630 75650 75680 75700 75710 75730 75750 75750 75700 75580 75430 75250 75070 74860 74660 74470 74300 74160 C-2

GE-NE-0000-0002-8817-01-R2 APPENDIX D TABLES FOR LONG-TERM CONTAINMENT RESPONSE TO DBA-LOCA FOR INPUT TO NPSH (K=147, SWT=90 OF)

Table Title Page D-1 One-Day Long-Term (600 Seconds to 1 Day) Response with D-2 RHR Heat Exchanger K-Value of 147 Btulsec-F and 90 F Service Water Temperature D-2 Twelve-Day Long-Term (1 to 12 Days) Response with RHR D-5 Heat Exchanger K-Value of 147 Btulsec-F and 90 F Service Water Temperature D-1

GE-NE-0000-0002-88 17-01 -R2 Table D-1 One-Day Long-Term (600 Seconds to 1 Day) Response with RHR Heat Exchanger K-Value of 147 Btu/sec-0F and 90 0F Service Water Temperature TIME (Seconds) 0.0 591.3 875.9 1223.9 1606.1 2010.1 2411.6 2807.4 3210.8 3608.6 4003.4 4393.7 4777.2 5179.0 5570.1 5970.0 6367.7 6768.1 7161.3 7564.3 7970.1 8368.0 8764.4 9165.6 9568.6 9972.0 10374 10778 11579 12374 13176 13976 14777 15579 16380 Wetwell Pressure with Leakage (PSIA) 14.26 23.11 17.89 17.84 17.79 17.74 17.86 18.05 18.23 18.45 18.62 18.76 18.97 19.14 19.32 19.54 19.63 19.76 19.88 20.00 20.10 20.20 20.29 20.38 20.46 20.54 20.61 20.69 20.84 20.98 21.13 21.24 21.32 21.41 21.48 Wetwell Pressure w/o Leakage (PSIA) 14.26 23.11 17.89 17.84 17.79 17.75 17.86 18.06 18.23 18.46 18.63 18.77 18.98 19.16 19.33 19.55 19.64 19.78 19.90 20.01 20.12 20.22 20.31 20.39 20.48 20.56 20.64 20.71 20.86 21.00 21.16 21.27 21.35 21.44 21.51 Suppression Pool Temperature (F) 90.0 149.1 155.1 158.5 161.1 163.4 165.2 166.8 168.3 169.8 171.2 172.4 173.6 174.7 175.8 176.8 177.8 178.7 179.5 180.3 181.1 181.8 182.4 183.1 183.7 184.3 184.8 185.4 186.3 187.2 188.0 188.8 189.5 190.1 190.6 Suppression Pool Volume (CU FT) 68000 71040 75510 76260 76320 76220 75770 75400 75050 74850 74710 74580 74500 74440 74340 74300 74290 74270 74260 74250 74240 74230 74230 74230 74240 74240 74240 74250 74260 74270 74250 74270 74270 74270 74270 D-2

GE-NE-0000-0002-8817-01 -R2 TIME (Seconds) 17184 17981 18787 19589 20383 21191 21992 22790 23594 24390 25191 25985 26784 27584 28389 29190 29987 30783 31582 31623*

32381 33177 33973 34773 35571 36375 37175 37972 38773 39579 40382 41185 41984 42780 43572 44366 45163 45968 46777 47580 48382 49190 49988 Wetwell Pressure with Leakage (PSIA) 21.54 21.60 21.65 21.72 21.73 21.76 21.78 21.80 21.82 21.84 21.85 21.86 21.87 21.88 21.88 21.89 21.89 21.89 21.89 21.89 21.88 21.88 21.87 21.86 21.85 21.84 21.83 21.82 21.81 21.80 21.78 21.77 21.75 21.74 21.72 21.71 21.69 21.67 21.65 21.62 21.59 21.55 21.53 Wetwell Pressure w/o Leakage (PSIA) 21.58 21.63 21.68 21.75 21.77 21.80 21.83 21.85 21.87 21.89 21.90 21.92 21.93 21.94 21.94 21.95 21.95 21.95 21.95 21.95 21.95 21.94 21.94 21.93 21.92 21.92 21.91 21.90 21.89 21.87 21.86 21.85 21.84 21.82 21.81 21.79 21.78 21.76 21.74 21.71 21.68 21.65 21.62 Suppression Pool Temperature (F) 191.1 191.5 191.9 192.3 192.6 192.9 193.1 193.3 193.5 193.6 193.7 193.9 193.9 194.0 194.1 194.1 194.1 194.1 194.2 194.2 194.2 194.1 194.1 194.1 194.0 194.0 193.9 193.9 193.8 193.7 193.6 193.5 193.4 193.2 193.1 193.0 192.9 192.7 192.6 192.4 192.3 192.1 192.0 Suppression Pool Volume (CU FT) 74270 74270 74200 74240 74240 74240 74240 74230 74230 74210 74210 74190 74180 74150 74140 74110 74110 74090 74060 74060 74040 74020 73990 73970 73940 73900 73880 73840 73810 73770 73730 73700 73670 73640 73610 73580 73540 73520 73480 73450 73430 73390 73360 D-3

GE-NE-0000-0002-8817-01 -R2 TIME (Seconds) 51589 53198 54800 56403 57998 59601 61195 62794 64391 65986 67588 69180 70776 72372 73962 75555 77152 78738 80327 81917 83501 85094 86683 87000 Wetwell Pressure with Leakage (PSIA) 21.46 21.39 21.32 21.27 21.22 21.16 21.08 21.00 20.91 20.83 20.75 20.67 20.59 20.51 20.43 20.36 20.28 20.21 20.14 20.07 20.00 19.92 19.85 19.84 Wetwell Pressure w/o Leakage (PSIA) 21.56 21.49 21.43 21.38 21.33 21.27 21.20 21.12 21.04 20.96 20.88 20.80 20.73 20.65 20.58 20.50 20.43 20.36 20.29 20.22 20.16 20.09 20.02 20.00 Suppression Pool Temperature (9 )

191.7 191.3 191.0 190.7 190.5 190.3 189.9 189.5 189.0 188.6 188.2 187.8 187.3 186.9 186.5 186.1 185.7 185.3 184.9 184.5 184.1 183.7 183.3 183.2 Suppression Pool Volume (CU FT) 73290 73230 73200 73190 73170 73070 72970 72900 72820 72770 72710 72640 72590 72540 72470 72410 72350 72300 72230 72180 72120 72070 72020 72010

Time of Peak Suppression Pool Temperature D-4

GE-NE-0000-0002-8817-01-R2 Table D-2 Twelve-Day Long-Term (1 to 12 Days) Response with RHR Heat Exchanger K-Value of 147 Btu/sec-F and 90 0p Service Water Temperature TIME (Seconds) 0.0 85013 96925 108836 120766 132690 144598 156508 168405 180314 192210 204123 216042 227971 239934 251899 263862 275838 287820 299816 311812 323807 335815 347826 359822 371837 383825 395851 407879 419910 431954 443978 456030 468055 480095 492139 504214 Wetwell Wetwell Suppression Suppression Pressure Pressure w/o Pool Pool Volume with Leakage Leakage Temperature (CU F (PSIA)

(PSIA)

(F) 14.26 14.26 90.0 68000 19.93 20.09 183.7 72080 19.42 19.61 180.8 71690 18.96 19.16 178.0 71360 18.57 18.79 175.6 71100 18.23 18.48 173.4 70850 17.94 18.20 171.4 70650 17.67 17.96 169.6 70450 17.43 17.74 167.9 70260 17.20 17.53 166.2 70080 17.01 17.35 164.6 69920 16.79 17.16 163.0 69770 16.60 16.99 161.6 69640 16.44 16.85 160.4 69540 16.30 16.73 159.4 69440 16.17 16.62 158.4 69340 16.05 16.52 157.5 69260 15.94 16.43 156.7 69180 15.83 16.34 155.9 69100 15.73 16.26 155.1 69030 15.63 16.18 154.4 68960 15.53 16.10 153.6 68880 15.43 16.02 152.9 68820 15.33 15.94 152.1 68750 15.24 15.87 151.4 68680 15.15 15.80 150.6 68620 15.05 15.73 149.9 68560 14.96 15.66 149.2 68500 14.88 15.59 148.5 68440 14.80 15.53 147.9 68390 14.74 15.49 147.4 68350 14.68 15.45 147.0 68320 14.63 15.42 146.6 68280 14.57 15.39 146.2 68240 14.52 15.36 145.9 68210 14.48 15.33 145.6 68180 14.43 15.30 145.3 68150 D-5

GE-NE-0000-0002-881 7-01 -R2 TIME (Seconds) 516267 528322 540385 552454 564515 576588 588643 600710 612773 624843 636905 648974 661033 673076 684190 694142 704024 713852 723670 733497 743309 753101 762884 772653 782441 792188 801969 811741 821513 831278 841037 850808 860552 870316 880099 889834 899549 909281 919004 928736 938472 948209 957928 Wetwell Pressure with Leakage (PSIA) 14.38 14.34 14.29 14.25 14.20 14.16 14.12 14.07 14.03 13.98 13.94 13.90 13.86 13.81 13.77 13.74 13.71 13.67 13.64 13.60 13.58 13.56 13.53 13.50 13.47 13.43 13.40 13.37 13.33 13.30 13.27 13.23 13.20 13.17 13.13 13.10 13.07 13.03 13.00 12.97 12.93 12.90 12.87 Wetwell Suppression Suppression Pressure w/o Pool Pool Volume Leakage Temperature (CU Y)

(PSIA)

(F) 15.27 145.0 68120 15.25 144.7 68090 15.22 144.5 68060 15.20 144.2 68030 15.18 143.9 68000 15.15 143.6 67970 15.13 143.3 67940 15.10 143.1 67920 15.08 142.8 67880 15.06 142.5 67860 15.03 142.2 67830 15.01 142.0 67800 14.99 141.7 67780 14.97 141.4 67750 14.94 141.2 67730 14.93 140.9 67710 14.91 140.7 67680 14.89 140.5 67670 14.87 140.3 67650 14.86 140.0 67630 14.85 139.8 67610 14.85 139.6 67590 14.83 139.4 67570 14.82 139.2 67560 14.80 139.0 67540 14.78 138.8 67520 14.76 138.6 67500 14.75 138.3 67480 14.73 138.1 67470 14.71 137.9 67450 14.69 137.7 67430 14.68 137.5 67410 14.66 137.2 67400 14.64 137.0 67390 14.62 136.8 67370 14.61 136.6 67350 14.59 136.4 67340 14.57 136.1 67320 14.55 135.9 67300 14.54 135.7 67290 14.52 135.5 67280 14.50 135.2 67260 14.49 135.0 67240 D-6

GE-NE-0000-0002-8817-01 -R2 Wetwell Wetwell Suppression Suppression Pressure Pressure w/o Pool Puppresion TIME (Seconds) 967651 977376 987101 996801 1006513 1016193 1025880 1035557 1045366 1055177 1064919 1074656 1084339 1093969 1100000 with Leakage Leakage (PSUI) 12.84 12.81 12.77 12.74 12.71 12.68 12.65 12.63 12.59 12.54 12.50 12.48 12.44 12.41 12.40 (PSIA) 14.47 14.45 14.44 14.42 14.41 14.39 14.38 14.37 14.35 14.32 14.29 14.29 14.26 14.24 14.25 Temperature 14) 134.8 134.6 134.4 134.1 133.9 133.7 133.6 133.5 132.8 132.4 132.1 131.9 131.7 131.6 131.5 (CU FT) 67230 67220 67200 67190 67180 67160 67150 67150 67180 67100 67050 67200 67040 66960 67030 D-7

GE-NE-0000-0002-8817-1 -R2 APPENDIX E TABLES FOR LONG-TERM CONTAINMENT RESPONSE TO DBA-LOCA FOR INPUT TO NPSH (K=147, SWT=94 OF)

Table Title Page E-1 One-Day Long-Term (600 Seconds to 1 Day) Response with E-2 RHR Heat Exchanger K-Value of 147 Btusec-F and 94 F Service Water Temperature E-2 Twelve-Day Long-Term (1 to 12 Days) Response with RHR E-5 Heat Exchanger K-Value of 147 Btulsec-0F and 94 VF Service Water Temperature E-1

GE-NE-0000-0002-8 17-01 -R2 E-2

GE-NE-000-0002-8817-01-R2 Table E-1 One-Day Long-Term (600 Seconds to 1 Day) Response with RHR Heat Exchanger K-Value of 147 Btu/sec-°F and 94 TF Service Water Temperature TIME (Seconds) 0.0 5913 871.1 1208.6 1583.4 1986.6 2387.9 2786.2 3190.9 3594.9 3994.9 4396.3 4771.6 5167.5 5563.7 5971.7 6369.3 6760.6 7156.4 7554.9 7951.7 8352.9 8751.3 9153.3 9555.3 9955.4 10350 10751 11550 12345 13145 13937 14727 15524 16320 Wetweil Pressure with Leakage (PSIA) 14.26 23.11 17.94 17.91 17.86 17.82 17.93 18.13 18.32 18.56 18.73 18.92 19.07 19.27 19A5 19.70 19.79 19.93 20.06 20.19 20.30 20.40 20.50 20.59 20.68 20.77 20.85 20.93 21.08 21.23 21.42 21.51 21.62 21.72 21.81 Wetwell Suppression Suppression Pressure wlo Pool Pnnp Vrsion Leakage (PSLA) 1426 23.11 17.94 17.91 17.86 17.83 17.93 18.14 18.32 18.57 18.74 18.93 19.08 19.28 19.46 19.71 19.81 19.95 20.08 20.20 2031 20.42 20.52 20.61 20.70 20.79 20.87 20.95 21.11 21.26 21.44 21.54 21.65 21.75 21.84 Temperature (E) 90.0 149.1 155.0 158.4 161.0 163A 165.2 166.9 168.5 170.0 171.5 172.8 174.0 175.2 176.3 177.4 178.4 179.3 180.2 181.0 181.8 182.5 183.2 183.9 184.5 185.2 185.7 186.3 187.3 188.3 189.2 190.0 190.7 191.3 191.9 (CU I V f)

(CUMF 68000 71040 75420 76260 76330 76220 75770 75400 75040 74850 74700 74590 74510 74450 74390 74310 74310 74290 74280 74270 74260 74250 74260 74260 74270 74270 74280 74300 74310 74330 74320 74350 74360 74350 74360 E-3

GE-NE-0000-0002-8817-01-R2 TIME (Seconds) 17115 17914 18716 19520 20316 21114 21912 22706 23504 24297 25097 25895 26690 27486 28281 29085 29889 30692 31494 32292 32422*

33099 33898 34696 35486 36279 37076 37865 38664 39468 40262 41063 41857 42656 43452 44252 45044 45841 46639 47431 48235 49033 49830 Wetwell Pressure with Leakage (PSIA) 21.88 21.94 22.00 22.04 22.12 22.15 22.18 22.21 22.23 22.26 22.28 22.30 22.32 22.33 22.34 22.35 22.35 22.35 22.35 22.35 22.35 22.35 22.34 22.34 22.33 22.32 22.32 22.30 22.29 22.28 22.27 22.26 22.25 22.23 22.22 22.21 22.19 22.18 22.16 22.14 22.12 22.10 22.07 Wetwell Suppression Suppression Pressure w/o Pool Sppression Leakage (PSIA) 21.91 21.98 22.04 22.08 22.16 22.19 22.22 22.25 22.28 22.31 22.33 22.35 22.37 22.38 22.39 22.40 22.41 22.41 22.42 22.42 22.42 22.42 22.41 22.41 22.40 22.39 22.39 22.38 22.37 22.36 22.35 22.34 22.33 22.32 22.31 22.29 22.28 22.27 22.25 22.24 22.22 22.20 22.17 Temperature (F) 192.5 192.9 193.4 193.8 194.2 194.5 194.8 195.0 195.2 195.4 195.6 195.7 195.9 196.0 196.0 196.1 196.2 196.2 196.2 196.2 196.2 196.2 196.2 196.2 196.2 196.2 196.1 196.1 196.0 195.9 195.9 195.8 195.7 195.6 195.5 195.4 195.3 195.2 195.0 194.9 194.8 194.7 194.5 U.

Iv V -

(CU FT) 74360 74360 74370 74290 74330 74350 74350 74350 74350 74340 74340 74320 74310 74290 74280 74270 74240 74220 74210 74180 74180 74160 74150 74120 74090 74070 74030 74000 73970 73940 73910 73870 73840 73810 73780 73760 73730 73700 73660 73630 73610 73580 73540 E-4

GE-NE-0000-0002-8817-0 1 -R2 TIME (Seconds) 51425 53036 54636 56246 57846 59433 61024 62617 64213 65806 67396 68987 70574 72165 73752 75341 76939 78529 80120 81710 83301 84897 86492 87000 Wetwell Wetwell Suppression Pressure Pressure w/o Pool Suppression with Leakage Leakage Temperature (CU lm (PSIA)

(PSIA)

(F) 22.01 22.11 194.3 73490 21.95 22.05 194.0 73420 21.88 21.99 193.6 73360 21.81 21.92 193.3 73290 21.74 21.85 193.0 73240 21.67 21.79 192.7 73220 21.62 21.74 192.4 73210 21.57 21.69 192.2 73170 21.50 21.63 191.9 73060 21.43 21.55 191.6 72980 21.34 21.47 191.2 72910 21.26 21.39 190.8 72850 21.18 21.32 190.4 72790 21.11 21.25 190.0 72740 21.03 21.17 189.6 72680 20.95 21.10 189.2 72620 20.88 21.03 188.8 72570 20.81 20.96 188.4 72510 20.73 20.88 188.0 72450 20.65 20.81 187.6 72390 20.58 20.74 187.3 72340 20.51 20.67 186.9 72280 20.44 20.60 186.5 72230 20.42 20.58 186.4 72200

Time of Peak Suppression Pool Temperature E-5

GE-NE-0000-0002-8817-01 -R2 Table E-2 Twelve-Day Long-Term (1 to 12 Days) Response with RHR Heat Exchanger K-Value of 147 Btu/sec-0 F and 94 TF Service Water Temperature TIME (Seconds) 0.0 84817 96735 108615 120459 132330 144180 156013 167850 179661 191489 203344 215199 227040 238901 250764 262666 274548 286448 298349 310280 322204 334115 346058 357997 369957 381907 393845 405793 417742 429706 441677 453683 465648 477615 Wetwell Pressure with Leakage (PSIA) 14.26 20.51 20.00 19.54 19.15 18.81 18.50 18.22 17.97 17.73 17.50 17.30 17.11 16.94 16.79 16.65 16.52 16.40 16.29 16.18 16.07 15.96 15.86 15.76 15.66 15.56 15.47 15.37 15.28 15.20 15.13 15.06 15.01 14.95 14.90 Wetwell Pressure without Leakage (PSIA) 14.26 20.67 20.19 19.74 19.37 19.05 18.77 18.51 18.27 18.06 17.85 17.67 17.49 17.34 17.21 17.10 16.99 16.89 16.80 16.71 16.62 16.54 16.45 16.37 16.29 16.21 16.14 16.06 15.99 15.93 15.88 15.83 15.80 15.76 15.73 Suppression Pool Suppression Pool Volume Temperature (CU MiT)

(F) 90.0 68000 186.9 72290 184.1 71890 181.5 71560 179.2 71280 177.2 71040 175.3 70830 173.5 70630 171.8 70430 170.2 70250 168.6 70080 167.0 69910 165.6 69790 164.5 69670 163.4 69560 162.5 69470 161.6 69380 160.8 69290 160.0 69210 159.2 69140 158.5 69060 157.7 68980 157.0 68900 156.2 68840 155.5 68770 154.8 68700 154.0 68640 153.3 68580 152.6 68510 152.0 68460 151.5 68420 151.0 68380 150.7 68340 150.3 68300 150.0 68270 E-6

GE-NE-0000-0002-8817-01-R2 Wetwell Wetwell Suppression TIME Pressure Pressure Pool Suppression with without Pool Volume (Seconds)

Leakage Leakage (CUm (PSIA)

(PSIA) 489576 14.85 15.70 149.7 68230 501509 14.80 15.67 149.4 68190 513499 14.75 15.64 149.1 68160 525485 14.71 15.62 148.8 68130 537476 14.66 15.59 148.5 68100 549455 14.61 15.56 148.2 68070 561455 14.57 15.54 148.0 68040 573455 14.52 15.51 147.7 68010 585433 14.47 15.48 147.4 67970 597425 14.43 15.46 147.1 67950 609406 14.38 15.43 146.9 67920 621398 14.34 15.41 146.6 67890 633347 14.29 15.38 146.3 67860 644665 14.25 15.36 146.1 67830 654698 14.21 15.34 145.8 67810 664494 14.18 15.32 145.6 67790 674245 14.14 15.30 145.4 67760 684006 14.11 15.28 145.2 67740 693765 14.07 15.26 144.9 67720 703547 14.04 15.24 144.7 67700 713274 14.00 15.22 144.5 67680 723047 13.97 15.20 144.3 67660 732808 13.93 15.19 144.1 67640 742558 13.90 15.17 143.8 67620 752305 13.86 15.15 143.6 67600 762028 13.83 15.13 143.4 67580 771758 13.79 15.11 143.2 67560 781475 13.77 15.10 143.0 67540 791186 13.74 15.10 142.7 67520 800896 13.72 15.09 142.5 67510 810600 13.69 15.07 142.3 67490 820310 13.65 15.05 142.1 67470 829994 13.62 15.03 141.9 67460 839696 13.58 15.01 141.7 67440 849428 13.55 15.00 141.5 67420 859157 13.52 14.98 141.3 67400 868864 13.48 14.96 141.1 67380 878568 13.45 14.94 140.8 67360 888287 13.41 14.92 140.6 67350 898018 13.38 14.90 140.4 67330 907726 13.34 14.89 140.2 67310 917416 13.31 14.87 140.0 67300 E-7

GE-NE-0000-0002-8817-01 -R2 TIME (Seconds) 927111 936799 946487 956176 965872 975534 985230 994904 1004609 1014311 1023981 1033678 1043356 1053033 1062711 1072379 1082070 1091757 1100000 Wetwell Pressure with Leakage (PSIA) 13.28 13.24 13.21 13.18 13.14 13.11 13.07 13.04 13.01 12.98 12.95 12.92 12.90 12.87 12.85 12.82 12.80 12.78 12.76 Wetwell Pressure without Leakage (PSIA) 14.85 14.83 14.81 14.79 14.78 14.76 14.74 14.72 14.71 14.69 14.68 14.67 14.66 14.65 14.64 14.63 14.63 14.62 14.61 Suppression Pool Suppression Pool Volume Temperature (CU IT (F) 139.7 67280 139.5 67260 139.3 67260 139.1 67240 138.9 67230 138.6 67210 138.4 67190 138.2 67180 138.0 67160 137.8 67150 137.6 67140 137.5 67130 137.4 67120 137.3 67110 137.2 67100 137.1 67080 137.0 67080 136.9 67070 136.8 67060 E-8

GE-NE-0000-0002-88 17-01-R2 APPENDIX F PLOTS FROM SAFER/GESTR RUNS Figure Title Page F.1-SAFER Results for Isolation Event, 102% of 1775 MWT, ANS 5.1 + 2ca D.H 1 Water Level In Hot and Average Channels.

F-2 2 Water Level in Upper Plenum and Bypass F-3 3 Water Level in Regions 6 and 7 and Downcomer.

F-4 4 Reactor Vessel Pressure.

F-5 5 SRV, ADS and Break Flows.

F-6 6 ECCS Flows.

F-7 F.2-SAFER Results for 0.01 ft2 Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2cr D.H.

1 Water Level in Hot and Average Channels.

F-8 2 Water Level in Upper Plenum and Bypass F-9 3 Water Level in Regions 6 and 7 and Downcomer.

F-10 4 Reactor Vessel Pressure.

F-11 5 SRV, ADS and Break Flows.

F-12 6 ECCS Flows.

F-13 F.3-SAFER Results for 0.10 ft2 Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2cr D.H.

1 Water Level in Hot and Average Channels.

F-14 2 Water Level in Upper Plenum and Bypass F-15 3 Water Level in Regions 6 and 7 and Downcomer.

F-16 4 Reactor Vessel Pressure.

F-17 5 SRV, ADS and Break Flows.

F-18 6 ECCS Flows.

F-19 F-1

MONTICELLO 1 HOT CHANEL 2 AVERAGE CHANNEL E TOP OF ACTIVE FL I & 2 are the same except hot channel

-L is uncovered slightly less BATTERY FAILURE 60.

1 4

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h-tI U-

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t I

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_~~~~~~~~~~~~~~~~~~~~~~~~~~~~

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0 h6TAIRR OOd 22d11 i B16 S.7 0.8 TIME 1.6 (SECONDS) 2.4

3. 2X' 0' Figure F. I -I Water Level in Hot and Average Channels.

SAFER Results for Isolation Event, 102% of 1775 MWT, ANS 5.1 + 2a D.H.

60.

40.

w P

tl 0

0 00 00 0

bf 0-

-J Lii 20.

O.

0.

MAR 0R 5 2M32O11 1818.7 0.8 1.6 TIME (SECONDS) 2.4 3.2%10" Figure F. 1-2 Water Level in Upper Plenum and Bypass.

SAFER Results for Isolation Event, 102% of 1775 MWT, ANS 5.1 + 2a D.H.

60.

40.

8~T tz 0

00 I?0 0

N)

~0 00

--4 6

i*

4 I-I LLJ c-J uJ CK 20.

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TAFR 00a185 22011W 1sle.7 0.8 1.6 2.4 3.2xl0 TIME (SECONDS)

Figure F. 1-3 Water Level in Regions 6 and 7 and Downcomer.

SAFER Results for Isolation Event, 102% of 1775 MWT, ANS 5.1 + 2a D.H.

MONTICELLO

VESSEL PRESSURE BATTERY FAILURE I.

X10 Q.

It1 2

£

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0TA01 15 22M 1 Il.7 0.8 TIME 1.6 (SECONDS) 2.4 3.2X10' Figure F. 1-4 Reactor Vessel Pressure.

SAFER Results for Isolation Event, 102% of 1775 MWT, ANS 5.1 + 2a D.H.

1.5 xlO' 1.

Iq1

-J Ci 0z 0

0 0t' 00 00 0

0.5 L]

C3i L.L 0.

WSTARR 00affi 0.8 I.E TIME (SECONDS) 3 2.A

3. 2X10' Figure F.1-5 SRV, ADS and Break Flows.

SAFER Results for Isolation Event, 102% of 1775 MWT, ANS 5.1 + 2a D.H.

MONTICELLO BATTERY FAILURE I IPCI

LPCS #1 LPCS #2 4

1 LPCI IN 9 PrT TN 1 LOOP 1

o

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0 I-

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00 bA I.-A 0.5 O.

0.

TNM WM&5 2003I1 1818.7 0.8 TIME 1.6 (SECONDS) 2,4 3.2X108 Figure F.1-6 ECCS Flows.

SAFER Results for Isolation Event, 102% of 1775 MWT, ANS 5.1 + 2a D.H.

I & 2 are the same except hot channel 60.

40.

tri 0k 0

00~

00

~0 00 1-U-

-J Ld-20.

O. _

O.

WTAIRR 00092 20020314 1028.9 0.8 TIME 1.6 2.4 (SECONDS) 3.2X10' Figure F.2-1 Water Level in Hot and Average Channels.

SAFER Results for 0.01 ft2 Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2a D.H.

60.

40.

I" 000 0

00 00

-4

-j LI I--

20.

O.

MTnAM OeM2 2O02O!14 1.9 0.8 TIME 1.6 2.4 (SECONDS) 3.2x101 Figure F.2-2 Water Level in Upper Plenum and Bypass.

SAFER Results for 0.01 ft2 Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2a D.H.

BO.

40.4 O.-

0 0

tz 0

0 (D

0 0

0<D

~0 00 00-CD I-Li-IJ LL I-Q3 20.

0. -

O.

WSTAR 00M2 aMM14l t0M.9 0.8 1.E TIME (SECONDS) 6 2.4 3.2X10' Figure F.2-3 Water Level in Regions 6 and 7 and Downcomer.

SAFER Results for 0.01 ft2 Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2a D.H.

lid I-8 o0 0

00

-J S-

'-4 U) 0C 0.4 -

Lii i_

an (L

O. L 0.

WSTARM 0002 20205M 1028.9 0.8 1.6 2.4 3.2x10' TIME (SECONDS)

Figure F.2-4 Reactor Vessel Pressure.

SAFER Results for 0.01 f& Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2a D.H.

1.

'10*

1-4 C)

LLJ

-J 0.5 _

LU I

%-jO O.

0.

hSTAR 001U2 2D2D1 J

0tz0 0

0 00 0

(D 00 I-0.8 1.6 TIME (SECONDS) 2.4 3.2x10 Figure F.2-5 SRV, ADS and Break Flows.

SAFER Results for 0.01 ft2 Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2a D.H.

1.5 x1 0 1.

OJJ 0

i8

-J 0:

LIJ J'

0 0

0 00 0

0.

O.

0.

MTM OM2 2WO2O514 1028.9 0.8 TIME 1.6 2.4 (SECONDS) 3.2X10' Figure F.2-6 ECCS Flows.

SAFER Results for 0.01 ft2 Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2 D.H.

MONTICELLO BATTERY FAILURE i HOT CANNEL a AVERAGE CHANNEL 4 TOP OF ACTIVE FU I & 2 are the same except hot channel U is uncovered slightly ess 60.

40.

I

+

i

'r1 I-C-

LUJ I-20.

2 1

12 I

a I

a 1 I~~~~~~~~~~

Vessel filled solid, disregard data after 8 14 seconds.

-I I I I I I

I I I

__I_

00 0

00 CD 0

6 o

00

-J 0.I 0

WGO26OM 2O08l lw, 4

.0 0.8 TIME 1.6 (SECONDS) 2.4 3.2110' Figure F.3-1 Water Level in Hot and Average Channels.

SAFER Results for 0.10 ft2 Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2a D.H.

60.

40.

ITJ I

0 0

9 0

00 0

00

-J ILU

-j I

20.

O.

0.

fT O aME 20 Ii 1009.0 0.8 TIME 1.6 (SECONDS) 2.A 3.2'10 Figure F.3-2 Water Level in Upper Plenum and Bypass.

SAFER Results for 0.10 f 2 Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2a D.H.

60.

40.

'r1 P.-&

0 00 0tj3 00 o.

0I-

-1J uJ I-20.

O. _-

0.

WIRTA 02WE Of~~il I009.0 0.8 TIME 1.6 (SECONDS) 2.4 3.2X10' Figure F.3-3 Water Level in Regions 6 and 7 and Downcomer.

SAFER Results for 0.10 ft2 Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2a D.H.

MONTICELLO BATTERY FAILURE VESSEL PRESSURE 1.2 xi 0 0.8 I

I I.

4 4

4.

4 III

.4 I

to) 00

-4 0

'-q CO a_

O.A n

Of Lai 0-I Vessel filled solid, disregard data after 814 seconds

_ \\ I I

I I

1 1

I_

0.

HTWTN O2AE 0.8 1.6 TIME (SECONDS) 2.4 3.210 Figure F.3-4 Reactor Vessel Pressure.

SAFER Results for 0.10 ft2 Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2a D.H.

1.5 10' 1.

0 C)

LUJ

-J

%-S 0

0 0.5 Ld

-J Li-0.

0.

WRTARR 02s 2DO281i 1000.0 0.8 TIME 1.6 2.4 (SECONDS) 3.2x10' Figure F.3-5 SRV, ADS and Break Flows.

SAFER Results for 0.10 f Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2c D.H.

x10 I"

U 0.,5 _

co_

I

0.

-j IL LL_

0. _

0.

FMTA 02E 20020314 1009.0 0

0 t'J 00

-J b

0.8 1.6 2.A TIME (SECONDS) 3.2X10' Figure F.3-6 ECCS Flows.

SAFER Results for 0. 10 ft2 Rec. Ln. Brk., 102% of 1775 MWT, ANS 5.1 + 2a D.H.

GE-NE-0000-0002-88 17-0 1-R2 APPENDIX G PLOTS FROM SHEX RUNS FOR ISOLATION EVENTS, SMALL BREAKS, AND DBA-LOCA WITH DIRECT POOL COOLING (K= 143.1, SWT= 900F)

Figure Title Page G.1-Reactor Isolation with One RHR Loop (K=143.1 Btu/Sec-'F, SWT = 901F)

1. Suppression Pool Temperature G-2

2. Drywell and Wetwell Pressure G-3

3. Reactor Pressure Vessel Pressure G-4 G.2-Liquid Line Break 0.01 Fe with One RHR Loop (K=143.1 BtulSec-F, SWT = 90°F)

1. Suppression Pool Temperature

-S5

2. Drywell and Wetwell Pressure G-6

3. Reactor Pressure Vessel Pressure G-7 G.3-Liquid Line Break 0.1 Fe with One RHR Loop (K=143.1 BtulSec-'F, SWT = 900F)

1. Suppression Pool Temperature G-8

2. Drywell and Wetwvell Pressure G-9

3. Reactor Pressure Vessel Pressure G-10 G.4-DBA-LOCA - Loss of Diesel Generator (K=143.1 BtulSec-F, SWT = 90°F)

1. Suppression Pool G-11

2. Drywell and Wetwell Pressure G-12

3. Reactor Pressure Vessel Temperature G-13 G.5-Reactor Isolation with Two REIR Loops (K=143.1 Btu/Sec-F, SWT = 900EF)

1. Suppression Pool Temperature G-14

2. Drywell and Wetwvell Pressure G-15

3. Reactor Pressure Vessel Pressure G-16 G-1

GE-NE-0000-0002-8817-01-R2 LL C.D LU Ld L

LU O. I I"I SAYLES 26EOOCC 02302 1321.9

2.

3.

LOG TIME - SEC 5.

Figure G.1-1 Suppression Pool Temperature - Reactor Isolation with One RHR Loop (K=143.1 BaWSec-F, SWT= 90F)

G-2

GE-NE-0000-0002-8817-01 -R2 60.

40.

En Qua 0:

20 O.

l 1.

SAYLES 2EOOC9 O2MO2 1321.9

2.

3.

4.

5.

LOG TIME -

SEC Figure G.1-2 Drywell and Wetwell Pressure ReactorIsolation with One RH) Loop (=143.1 Btu/SecF, SWT= 9r)

(Since vessel surface heat transfer to the dryweil was not modeled, the containment pressure response Is underestimated)

G-3

GE-NE-0000-0002-8817-01-R2 MONTICELLO 0.0001 SO. FT SLB 1

V PRESSURE I.

-1o 5

5 I

I I I I_

TI u)

In cn UjL Of a) 0.

0 I1.

sianEs 2%OOC8C 02302 1321.9 2.

LOG TIME 3.

SEC i4.

5.

Figure G.1-3 Reactor Pressure Vesse Pressure - Reactor Isolation with One RHR Loop (K=143.1BtuISec-FSWT = 90F)

G4

GE-NE-0000-0002-8817-01-R2 MONTICELLO 0.01

0. FT SB I

SP TEMP 300.

t I

I I

200.

CD I-Uj a-a-_~~~~~~~~~

_,,,, I

_~~~~~~~~~~~~~~~~~~~~~~~~~~~

O.I 1.

SAYtES 2EOOc oM2y2 1!28.7

2.

3.

LOG TIME -

SEC 5.

Figure G.2-1 Sppression Pool Temperature - Liquid Llne Break 0.01 F# with One MRfR Loop (=143.1 Btu/Sec-F, SWT= 907)

G-5

GE-NE-0000-0002-8817-01 -R2 60.

40.

U)

A_

I LL U)

LU CL 20.

O.

l 1.

SAYLES 2iEOOCS5 Ma22 1328.7

2.

3T.

E -

5.

LOG TIME -

SEC Figure G.2-2 DryweU and Wetwel Pressure - Liquid Line Break.01 FtI with One RUR Loop (KC143.1 Btl/SecF, SWT = 90F)

G-6

GE-NE-0000-0002-8817-01-R2 c

a-En LI U0 a)a:

1.

SAYLES 2mooct M23O2 1!28.7

2.

3.

4.

5.

LOG TIME SEC Figure G.2-3 Reactor Pressure VesselPressure - LiquidLine Break 0.01 Ott with OneRHR Loop (KR143.1 Btu/Sec-F, SWT= 900F)

G-7

GE-NE-0000-0002-8817-01-R2 MONTICELLO 0.1 SO. FT SLB SP TEMP 500.

4-4-

200.

LL CD LLJ I-Ld Of LuJ C-100.

I

-,,, j I j I I j~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

A V.

I.

SAYLES 272007EI 0i2502 153.3

2.

3.

LOG TIME -

SEC A1.

5.

Figure G.3-1 Suppression Pool Temperature - Liquid Line Break al Ft2 with One RHR Loop (K-143.1 Btu/Sec-F, SWT - 90F)

G-8

GE-NE-0000-0002-88 17-01 -R2 60.

40.

I_

uLJ acn 20.

0.

I.

SAYLES 272001EI m202 1338.3

2.

3.

4.

5.

LOG TIME -

SEC Figure G.3-2 Drywell tnd Wetwell Pressure - Liquid Line Break 0.1 F9 witl One RHR Loop (K=143.1 Bt/Sec-F, SWT = 90" F)

G-9

GE-NE-0000-0002-8817-01 -R2

1. 5 I.

(nLI Ur)

LO 0*

0.51 0.

1.

SAYtES 272007E1 CA2302 15@3.3

2.

3.

LOG TIME SEC 5.

Figure G.3-3 Reactor Pressure Vessel Pressure - Liquid LIe Break 0. Ft2 will One RH Loop (K=143.1 BtWSeo-UF, SWT = 90)

G-10

GE-NE-0000-0002-8817-01-R2 MONT ICELLO CONT EgP TO l

SP TEFP OCA 300.

200.

I I

LL-CD uLJ I-LUJ LUJ 5-

~~~~~~~~~~~~~

I 100.

- I IIII O.I 1.

SAYLES 23OI53A 042302 1352.0

2.

3.

LOG TIME -

SEC 5.

Figure G.4-1 SppressiOn Pool Temperature - DBA-LOCA - Loss of Dlesel Generator (K143.1 Btu/Sec-F, SWPT 9G'F)

G-11

GE-NE-0000-0002-8817-01 -R2 60.

40.

C 1

20.

Ld U)

LU 0y O. I l I 1.

SAYaES 23C0153A OA2302 1552.0

2.

3.

4.

5.

LOG TIME -

SEC Figure G.4-2 Drywell and WetwellPressure - DBA-LOCA - Loss of Diesel Generator (K=143.l BtuSec-F, SWT-9OF)

G-12

GE-NE-0000-0002-88 17-01 -R2 Monticelo Containment Respone DBA-LOCA Loss of Diesel Power I

I5 U1 U0 U0 I

3W0 300.

250 200 150 100 an_

I I_

I I

0 I__

I 10 100 1000 ime eo

)

100000 Figure G.4-3 ReactorPressure Vessel Temperature - DBA-LOCA -LossofDiesel Generaor(K143.1Sec-FSWT90F)

G-13

GE-NE-0000-0002-8817-01-R2 MONT ICELLO 0.0001 SO. FT SLB Vs I

SP TEMP 2 DI 300.

200.

LL co I-L+/-J 0y, w)

Uf 100.

- I It I I I 0.

I.

SAYLES 290005A os0902 1o3.o

2.

3.

LOG TIME - SEC 5.

Figure G.S-I Suppression Pool Temperature - ReactorIsolon wth TwoAR Loops (K=143.1 B/aSe-"F SWT= 91F)

G-14

GE-NE-0000-0002-88 17-01 -R2 MONTICELLO 0.0001 SO. FT SB VS X OW PRESSURE W PRESSURE 20 60.

40.

LO cn LU 0LJ 20.

,,,,I I,_

II 0.

I.

SAYLES 290005iA 060902 MI03.0 2.

LOG 3.

TIME - SEC 5.

Figure G.5-2 Drywell and Wetwell Pressure - Reactor Isolation with Two RHR Loops (K=1I43.1 Btu/ Sec-F, SWT 901)

(Since vessel surface heat transfer to the drywell was not modeled, the containmentpressure response Is underestimated)

G-15

GE-NE-0000-0002-88 17-01 -R2 1.5 1.

U 0-I w

En cn U

Of wl 0.51

0. I i

1.

SAYLES 2900064A 050902 1H05.0 2.

LOG TIME 3.

SEC 5.

Figure G.5-3 Reactor Pressure Vessel Pressure - Reactor Isolation with Two RHR Loops (K-143.1 Btu/ Sec-F, SWT = 900)

G-16

GE-NE-0000-0002-88 17-01 -R2 APPENDIX H PLOTS FROM SHEX RUNS FOR SHORT-TERM NPSH DBA-LOCA Figure Title Page H-1 Suppression Pool Temperature - Short-Term NSPH DBA-H-2 LOCA H-2 Suppression Pool Volume - Short-Term NSPH DBA-LOCA H-3 H-3 Wetwell Temperature - Short-Term NSPH DBA-LOCA H-4 11-4 Wetwell Pressure with/without Containment Leakage -

11-5 Short-Term NSPH DBA-LOCA H-I

GE-NE-0000-0002-8817-01-R2 MONTICELLO CNT RESPONSE

- NWSH ST

' SP TEIP TO tOCA 30a 20 LL 0

LU r

I-

.0.

0.

SAtES 2m6ONC 06I2 oo0.0 150.

300.

LOG TIME - SEC 50.

600.

Figure H-l Suppression Pool Temperature - Short-Term NSPHDBA-LOCA H-2

GE-NE-0000-0002-8817-01-R2 Monaco Shoft.TVm NPSH I

I a

I

.II 0

Soo Thm Mconft 400 Figure 1-2 Suppression Pool Volume - Short-TermN SPHDBA-LOCA H-3

GE-NE-0000-0002-88 17-01 -R2 MONTICELLO CONT RESEO TO

- NWSH ST

, W AIRSPACE TEMP OCA 300.

4-

-4

+

20 Ll-(D LX l

I LU Ld I-0.

O. _,,

'0.-I 0.

SAILS 2wS09C 0S120 Uo.0 150.

300.

LOG TIME - SEC ASO.

600.

Figure 1-3 Wetweil Temperature - Slort-Term NSPHDBA-LOCA H-4

GE-NE-0000-0002-88 17-01-R2 MONTICELLO CONT RESPOW

- N'SH ST I AW PRSSUR WO PSS W

K TO 60.

+

4 0c 4O.

20.

a t

p

0.

I I

I a_

0.

9Se 2"0oIC utz oW.0 150.

500.

LOG TIME - SEC

50.

600.

Figure H-4 Wenell Pressure wfth/wthout Containment Leakage - Short-Term NSPHDBA-LOCA H-5

GE-NE-0000-0002-8817-01-R2 APPENDIX I PLOTS FROM SBEX RUNS FOR DBA-LOCA FOR LONG-TERM NPSH (K=147,SWT= 90 0F)

Figure Title Page 1.1-Long-Term (1 Day) DBA-LOCA for NPSH Evaluation (K=147 Btu/Sec-0F, SW Temp = 90F)

1. Suppression Pool Temperature 1-2

2. Suppression Pool Volume 1-3

3. Wetwell Temperature 1-4 4 Wetwell Pressure with/without Leakage I-5 1.2-Long-Term. (12 Day) DBA-LOCA for NPSH Evaluation (K=147 BtulSec-F, SW Temp = 90°F)

1. Suppression Pool Temperature I-6

2. Suppression Pool Volume 1-7

3. Wetwell Temperature I-8 4 Wetwell Pressure with/vithout Leakage 1-9 I-1

GE-NE-0000-0002-8817-01-R2 MONTICELLO COJT RESPNSE

- NWSH SP EMP TO 300.

200.

Lj-CD w

of 100.

CK ofi a-XJ LLJ i~~~~~~~~~~~~~~~~~~~~~~~

A_

-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

,,, t,,,,~~~~~~~~~~~~~~~~~~~~~~

O.I 1.

S&YLES 2Iorn 061202 o5.0

2.

3.

LOG TIME -

SEC A.

S.

FigureLil-l Suppression Pool TetperatureLong-Term (I Day)DBA-LOCA forNPSHEvawsalion (X=147Btu/Sec-'F, SWTemp 90"F)

I-2

GE-NE-0000-0002-8817-01-R2 Mdontleelo Long-Terem (1 Day) NPSH Suppression Pool Volume 10000o V

V I

z1A I

I I

7500 5000 SPLiq Vume 0

I 0

a_

2500 10 100 1000 Time (Seconds) 1000 1000 Figure L1-2 Suppression Pool Volume Long-Term (I Day) DBA-LOCA for NPSH Evaluation (=147BtuSec-!F, SW Temp 90F)

I-3

GE-NE-0000-0002-88 17-01 -R2 LiL.

CD w

ui I-

2.

3.

d.

5.

LOG TIME -

SEC sAms 2io172 061202 Cm.0 Figure Ll-3 WeiweU Temperature Long-Term (I Day) DBA-LOCA for NPSH Evaluaaon (K=147 Btu/Sec"F, SW Temp =9F)

I-4

GE-NE-0000-0002-8817-01-R2 MONTICELLO CONT RESPEE TO I

- NSH I 1 PSRE WO K

W PRESSURE W K

OCA 60.

AO.

20.

1,

,, I..

(O.

I.

SALES 2OI0?2 0B1202 OW.O

2.

3.

LOG TIME - SEC 5.

Figure 1.1-4 Wehwell Pressure wIh/wfthout Leakage Long-Term (1 Day) DBA-LOCAfor NPSHEvaluation (Kr147BtSec-F, SW Temp= 90F)

I-5

GE-NE-0000-0002-8817-01 -R2 300.

200.

MONT ICELLO CONT RESPONSE 147 12DA N'SH K X SP TEMP LL LO 0

LU I-C Ld i-

' ' ' ' I ' ' ' '

I_

100.

0.

2.

SAYLES 2W4I5O 0512o2 1716.3

3.

LOG TIME - SEC 5.

6.

Figure L2-1 Suppression Pool Temperature Long-Term (12 Day) DBA-LOCAfor NPSH Evaluation (K=147 Btu/Sec-F, SW Temp = 90(F)

I-6

GE-NE-0000-0002-8817-01-R2 Monticello Long-Term (12 Day) MPSH Suppresslon Pool Volume a

Sk E

I U

U0 aI soooo 50000 I-SP UquLd Volume 1000 10000 100000 Thm (89conds) 10000 10000000 Figure L2-2 Suppression Pool Volume Long-Term (12 Day) DBA-LOCAfor NPSHEvnluation (JK=147BftuSe'F, SW Temp - 90F)

I-7

GE-NE-0000-0002-8817-01 -R2 300.

200.

IL ED LUi I-Ld cz 100.

L^J L

I-O.

Il.

2.

SYAES 24401857 061285 1716.5

3.

A LOG TIME -

SEC 5.

6.

Figure L2-3 Weiwel Tenyperature Long-Term (12 Day) DBA-LOCAfor NPSH Evaluation (K=147 Bt/Sec-F, SW Temp = 90°F)

I-8

GE-NE-0000-0002-88 17-01-R2 MONTICELLO COWT ESPOEE d7 120A i

WH PRESSLRE W/O K I4W PRESSLRE W LK H Ki AAorl II l

40.

20.

I-'

Un 0C II LUJ 0E

-~~~ I I

I I I I I I I 2.

SC6LES tws1"7 cel2 171,.3

3.

LOG TIME - SEC 5.

6.

Figure 1.2-4 WeteliPressure withAv/thoutLeakage Long-Term (12 Day) DBA-LOCA for NPSHEvaluation (K=l47BtSeceF, SW Temp 90)

I-9

GE-NE-0000-0002-8817-01-R2 APPENDIX J PLOTS FROM SHEX RUNS FOR DBA-LOCA FOR LONG-TERM NPSH (K=147,SWT= 940F)

Figure Title Page J.1-Long-Term (1 Day) DBA-LOCA for NPSH Evaluation (K=147 Btu/Sec-°F, SW Temp = 94°F)

1. Suppression Pool Temperature J-2

2. Suppression Pool Volume J-3

3. Wetwell Temperature J-4 4 Wetwell Pressure with/without Leakage J-5 J.2-Long-Term (12 Day) DBA-LOCA for NPSH Evaluation (K=147 Btu/Sec-'F, SW Temp = 94°F)

1. Suppression Pool Temperature J-6

2. Suppression Pool Volume J-7

3. Wetwell Temperature J-8 4 Wetwell Pressure with/without Leakage J-9 J-1

GE-NE-0000-0002-8817-01-R2 MONTICELLO CONT RESPONSE 147 SlyI a

P TEMP N61 K 300.

200.

U-(9 LLJ Df 100.

D LX A-wd iii 11111 n

U.

1.

SAoUS MO2.1 o6I1K 1020.1

2.

3.

LOG TIME - SEC s.

Figure l.l-Suppression Pool TenmperatureLong-Tern (1 Day) DBA-LOCAforNPSHEW4uJaon* (KX147BIu/SecF, SWTenmp -94

)

J-2

GE-NE-0000-0002-88 17-01 -R2 Monticello Long-Tenn (1 Dey) NPSH Suppresson Pool Volume (SWT 94 F) 100000 75000 I

it 0-a.

=

25000 0

I

-SP Uquld Volume 10 100 1000 Tie (Semyndt) 10000 100000 Figure J.1-2 Suppression Pool Volume Long-Term ( Day) DBA-LOCA forNPSHEvaluation (-147Bft/Sec-F, SW Temp = 94f)

J-3

GE-NE-0000-0002-88 17-01 -R2 MONTICELLO CONT RES'cNE W

A7 SMM i W4 AIRSPACE TEP KI 300.

200.

Li Lli I

Of 100.

LU CL LUJ

-, I I I I I I I I 0.

1.

SAYLUS beMI uY140 1020.1

2.

5.

LOG TIME - SEC 5.

Figure J.1-3 WetweU Temperature Long-Term (I Day) DBA-LOCA for NPSH Evaluation (K(147Btu/Sec-F, SW Teip = 94F)

J-4

GE-NE-0000-0002-8817-01-R2 MONTICELLO WNT RESE NP 47 Bygi 1 I1 PRESSURE W/0 v WWa PRESSURE W 1 KI 60.

40.

_, I I I I,I, U,l U,

0I CY En 20.

0.

1.

SaUS 24I UeH2 1020.1

2.

3.

LOG TIME -

SEC 5.

Figure J.l-4 Wetweil Pressure with/wlthoutLeakage Long-Term (1 Day) DBA-LOCAfor NPSHEvaluatfon (K147Bfu/Sec-F, SW Temp = 94)

J-5

GE-NE-0000-0002-8817-01-R2 MONT ICELLO CONT REPOSE

/I sw9i is SP TEMP 300.

200.

LiL CD LU LUJ I=

,1 I I

I I I I I 100.

0.

2.

SAYLES 2W0ICM Oelim 2W21.0

3.

4.

LOG TIME -

SEC 5.

6.

Figure J.2-1 Suppression Pool Tenmperature Long-Term (12 Day) DBA-LOCAfor NPSH Evahialion (K=147BIuWSec-"F, SW Te9p = 94F)

J-6

GE-NE-0000-0002-8817-01 -R2 Monticello Long.Tenm (12 Day) NPSH Suppression Pool Volume (SWT a 94 F) 1.Ajuw I E

a I

W 7-4j 500

~

I -SP 25000 0

1000 10000 1I0000 Thw c

)

1000000 Figure J.2-2 Sppression Pool Volume Long-Term (12 Day) DBA-LOCAfor NPSHEvaluaion (K-147Bftu/Sec-F, SW Temp = 94F)

J-7

GE-NE-0000-0002-88 17-01 -R2 MONTICELLO CONT StEnSE W

A7 9

W4 AIRSPACE TEP H K1 300.

+

4 20A l

l l

3 LU (D

I-LUJ I

Of 100.

1 1I I

1 I I

I

_ _ _ _ __I_

0.

2.

SAYLES 24"O10o O60I2 2021.0

3.

A.

LOG TIME -

SEC 5.

6.

Figure J.2-3 Wedwei Temperature Long-Tenn (12 Day) DBA-LOCAfor NPSHEvaluation (K--147Btu/Se&"F, SW Temp = 94'F J-8

GE-NE-0000-0002-88 17-01 -R2 MONT ICELLO COT KSPNSE W

KI Al sw94 60.

40.

1 20.

O.

I

.I I I I I I I I

2.

3.

4.

5.

B.

Ceiac 202.0^;

LOG TIME - SEC Figure J.2-4 WetwellPressure with/withoutLeakage Long-Term (12Day) DBA-LOCAforNPSHEvaluaton (K=147Btu/SeccdF, SWT= 94F)

J-9

GE-NE-0000-0002-8817-01 -R2 APPENDIX K PLOTS FROM SHEX RUNS FOR DBA-LOCA WITH DIRECT POOL COOLING (K=147,SWT= 900F)

Figure Title Page DBA-LOCA - Loss of Diesel Generator RHR K=147 K

Btu/Sec-'F SW Temp = 90'F

1. Suppression Pool Temperature DBA-LOCA - Loss of K-2 Diesel Generator

2. Drywell and Wetwell Pressure DBA-LOCA - Loss of K-3 Diesel Generator

3. Drywell Temperature DBA-LOCA - Loss of Diesel K-4 Generator

4. Wetwell Temperature DBA-LOCA - Loss of Diesel K-5 Generator K-

GE-NE-0000-0002-8817-01-R2 30..

i20 S.

I 1 100

-SP TOmps I a

10 100 1000 10000 100000 Time (Seconds)

Figure K-i Suppression Pool Temperature DBA-LOCA wih Loss of Dksel Generator (K=147Btu/Sec-F. SWT= 9(P19 K-2

GE-NE-0000-0002-8817-01-R2 40 30 SIa 20 10 0

\\

{=~~~~~Wem Pl

+

4 10 100 1000 TV (SOCond.)

10000 100000 Figure K-2 Drywell and Wetwel PressureDBA-LOCA wth Loss of Desel Generator (K-147Bt Sec-'F, S T= 90F)

K-3

GE-NE-0000-0002-8817-01-R2 300 C 20 I

a s

l l

is I

-DfiM Teirn I

0 10 100 1000

!nm* (Second*)

10000 100000 Figure K-3 DryweU Temperature DBA-LOCA will Loss of Diesel Generator (K=147BWuSec-F, SWT = 9(tF)

K-4

GE-NE-0000-0002-8817-01 -R2 30 -

CM0 I

I-9 100-10 100 1000 1lo 1000 Th" (8-)

Figure K-4 Wetwell TemperatureDBA-LOCA with Loss of Dlesel Generator (Klb 47Bu/Sec'F, SPT= 90f;)

K-5

GE-NE-0000-0002-8817-01 -R2 APPENDIX L PLOTS FROM SHEX RUNS FOR DBA-LOCA WITH DIRECT POOL COOLING (K=147,SWT= 940F)

Figure Title Page DBA-LOCA - Loss of Diesel Generator RHR K=147 Btu/Sec-"F SW Temp = 94F

1. Suppression Pool Temperature DBA-LOCA - Loss of L-2 Diesel Generator

2. Drywell and Wetwell Pressure DBA-LOCA - Loss of L-3 Diesel Generator

3. Drywell Temperature DBA-LOCA - Loss of Diesel Lo Generator

4. Wetwell Temperature DBA-LOCA - Loss of Diesel L-S Generator L-1

GE-NE-0000-0002-8817-01-R2 30 c

T

-o20 I

Ii i lo CL 0

a.a.

0 0

0 1

-Sp TempeAt-o I

0 I

10 100 1000 10000 100000 Tfm. ("condo)

Figure L-1 Suppression Pool Tenperature DBA-LOCA will Loss ofdDleesel Geerator (K147BtuWSec-@F, SWT= 94F)

L-2

GE-NE-0000-0002-88 17-01-R2 30-20 L

10 I

0 m

Io lPf" I

0-10 1000 Time (soconds) 10000 100000 FigureL-2 Drywell and Wetwell PressureDBA-LOCA with Loss of Dlesel Generator (lKl147Bft/Sec-F, SOT= 94F)

L-3

GE-NE-0000-0002-8817-01-R2

-amu 200 c

I i100 4

4 I

-Dol Tenperbre l

U.'.

4 4

10 100 1000 Tih. (seconds) 10000 100000 Figure L-3 Dryweil TemperatureDBA-LOCA with Loss of Dlesel Generator (Kz=147Btu/SecF, SWT= 94F)

L-4

GE-NE-0000-0002-8817-01-R2 10 100 1000 10000 10000 rm. (

I)

Figure L-4 ewd Temperature DBA-LOC4 with Loss of Diesel Generator (K147 Bt/SeccF, SPT-= 94F)

L-5

v t e Letter Sequence Other
TAC:MB7185, (Open)
Initiation Request Acceptance... Administration Questions 19 August 2003 Answers 24 September 2003 Results Approval... Other: ML030170478, ML032721450, ML032970390, ML041410176, ML041540568
MONTHYEAR2003-08-31 [Table View]

ML032721450

Text

SUMMARY

SUMMARY

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